[en] Over a period of 4 and years the National Radioactive Waste Management Agency (Andra), working with a variety of Contractors mostly specializing in nuclear orientated mechanical applications, successfully designed, fabricated and demonstrated 2 very different prototype high level waste transport systems. The first system, based on air cushion technology, was developed primarily for very heavy loads (17 to 45 tonnes). The results of this work are described in a separate presentation (Paper 21) at this Conference. The second system, developed by Andra within the framework of the ESDRED Project, generally referred to as the 'Pushing Robot System' for vitrified waste canisters, is the subject of this paper. The 'Pushing Robot System' is a part of the French national disposal concept that is described in Andra's 'Dossier 2005'. The latter is a public document that can be viewed on Andra's web site (www.andra.fr). The 'Pushing Robot System' system is designed for the deep geological disposal (in clay formations) of 'C' type vitrified waste canisters. In its entirety the system provides for the transport, emplacement and, if necessary, the retrieval of those canisters. Nothing in the design of the Andra emplacement equipment would preclude its utilization in horizontal openings in other types of geological settings. Over a period of some 8 years Andra has developed the 'Pushing Robot System' in 3 phases. Initially there was only the 'Conceptual Design' (Phase 1) which was incorporated in the Dossier 2005. This was followed by Phase 2 i.e. the design and fabrication of a simplified full scale prototype system henceforth referred to a P1, which includes a Pushing Robot, a Dummy Canister and a Test Bench. P1 details were also incorporated in the Dossier 2005. Finally, during Phase 3, a second more comprehensive full scale prototype system P2 has been designed and is being assembled and tested this month. This system includes a Transport Shuttle, a Transfer Shielding Cask, a Docking Table, the Disposal Cell Mouth equipment and a full scale (100 m long) test bench, in addition to the Dummy Canister and a second generation Pushing Robot. The successful completion of the test campaign associated with the first prototype P1 confirmed the feasibility of emplacing 2 tonne/0.6 m diameter waste packages (canisters) containing long lived HLW in 40 m long horizontal bore holes (disposal cells) with only minimal annular clearance between the canister and the disposal cell liner. Preliminary testing of the second prototype P2 indicates that proper docking onto the cell mouth, followed by emplacement in 80 to 100 m long disposal cells, is possible. The developed technology is considered to be mature enough for a potential industrial application. The 2 prototypes (the 2nd and 3rd phases of the work) were executed within the framework of the ESDRED Project (Engineering Studies and Demonstration of Repository Designs) which is co-funded by the European Commission as part of the sixth EURATOM Research and Training Framework Programme (FP6) on nuclear energy (2002 - 2006). (author)

[en] Radioactive waste and spent nuclear fuel are generated in the Czech Republic as a consequence of the peaceful use of nuclear energy and ionising radiation in many industries, particularly in the generation of nuclear energy, health care (therapy, diagnostics), research, and agriculture. The current extent of utilisation of nuclear energy and ionising radiation in the Czech Republic is comparable with that of other developed countries. The Concept of Radioactive Waste and Spent Nuclear Fuel Management is a fundamental document formulating government and state authority strategy for the period up to approximately 2025 (affecting policy up to the end of the 21st century), concerning the organizations which generate radioactive waste and spent nuclear fuel. The Concept puts forward solutions to provide for the disposal of waste in compliance with requirements for the protection of human health and the environment without excessively transferring any of the current impacts of nuclear energy and ionising radiation utilisation to future generations. The Concept was approved by the government of the Czech Republic in 2002. According to the Concept high level waste and spent nuclear fuel generated at the Dukovany and Temelin nuclear power plants will eventually be disposed of in a deep geological repository. Such a repository should commence operation in 2065. Work aimed at selecting potentially suitable sites began in 1992, but the final site has not yet been determined. In compliance with the aforementioned Concept, the Radioactive Waste Repository Authority (RAWRA) is responsible for finding two suitable sites before 2015. The current stage of evaluation covers the whole of the Czech Republic and includes detailed criteria and requirements. Based on the latest findings RAWRA suggested six potential sites for further investigation at the beginning of 2003. (author)

[en] The concept for a disposal of high-level wastes (HLW) in Korea is based upon a multi barrier system composed of engineered barriers and its surrounding plutonic rock (Kang et. al., 2002). A repository is constructed in a bedrock of several hundred meters in depth below the ground surface. The engineered barrier system (EBS), which is similar to the configuration considered by many other countries, consists of the HLW-encapsulating disposal container, the buffer between the container and the wall of a borehole, and the backfill in the inside space of the emplacement room, to isolate the HLW from the surrounding rock masses. The engineering performance of a HLW repository is dependent, to a large extent, upon the thermal-hydro-mechanical (THM) behaviors in the buffer which are complicated by the processes such as the decay heat generated from the HLW, the ground water flowing in from the surrounding host rock, and the swelling pressure exerted by compacted bentonite. For this reason, the Korea Atomic Energy Research Institute (KAERI), to investigate the THM behaviors in the buffer of the Korean reference disposal system (KRS), planned large-scale tests to be conducted in two stages: a surface mock-up and then a full-scale 'in situ' test. This paper deals with the surface mock-up called as 'KENTEX' and presents the THM behaviors in the buffer which have been investigated from the KENTEX test. The KENTEX is a third scale of the KRS. It consists of five major components: a heating system, a confining cylinder, a hydration tank, bentonite blocks, and sensors and instruments. The heating system measures 0.41 m in diameter and 0.68 m in length, which includes three heating elements in its inside, capable of supplying a thermal power of 1 kW each. The confining cylinder, which plays a role of the wall of a borehole excavated in the host rock, is a steel body with a length of 1.36 m and an inner diameter of 0.75 m, the inside wall of which is lined with layers of geotextile and the outside wall of which is mounted wit h 24 nozzles with two metal filters inserted into the inside of each, to uniformly apply the groundwater to the outer surface of the bentonite blocks (i.e., hydration surface). The bentonite blocks are fabricated of 'Kyungju' bentonite which is being considered as a candidate buffer of the KRS. Total of 176 blocks are emplaced in 16 sections of the confining cylinder. The bentonite blocks have an average value of 13 % of a water content and the average dry density of the bentonite blocks in the confining cylinder is 1500 kg/m³. Total of sixty eight sensors are installed to measure the temperature, humidity (eventually, water content), and total pressure. And the heater control and data acquisition are operated automatically by means of a computer program. The KENTEX test which started on May 30, 2005 is now under a successful operation. The T-H-M behaviour in the bentonite blocks may allow us to draw preliminary and quantitative conclusions. The temperature reached a steady state in a short time after the te st start. The temperature was higher as it became closer to the heater, while it became lower as it was farther away from the heater. The water content had a higher value in a part close to the hydration surface than that in a heater part. The relative humidity data suggested that a hydration of the bentonite blocks might occur by different drying-wetting processes depending on their position. The total pressure was continuously increased by the evolution of the saturation front in the bentonite blocks and thereby the swelling pressure. There was also a contribution of a thermal expansion of the bentonite blocks near the heater and the capillary force in the dry bentonite blocks which the water did not reach from the hydration surface. (author)

[en] As for any other nuclear industrial facility, in a radioactive waste repository the various waste disposal operational activities from construction to closure can present a risk to human (workers and public) and the environment. In accordance with the December 30, 1991 French Waste Act, Andra has conducted feasibility studies regarding the disposal of HLW and ILW-LL waste in a clay host formation. The 'Dossier 2005 - Clay' includes a description of the operational safety analysis that was conducted for ILW-LL waste disposal in underground horizontal drifts. The objective of this paper is to present that safety analysis and its impact on the design at the feasibility stage. The safety analysis covered the operations from the reception of the waste transport casks to the disposal of the waste disposal package in its final emplacement location inside the disposal cell. Since the surface facilities' operations are similar to those of other nuclear ones, this paper focuses on the specificity of the deep repository, i.e. the operational safety and radioprotection aspects applied to the deep disposal drift. Andra has selected an ILW-LL design based on large horizontal drifts (diameters of 10 to 12 m, and lengths of 250 m). The primary waste packages are put inside a specific concrete overpack before their disposal. These overpacks are remotely stacked inside the horizontal drifts. The operational safety analysis aims to ensure that risks are kept under control through provisions in the design of the repository and by operating the facility in compliance with operational requirements and the safety functions. The requirements and the safety functions, developed at this stage of the feasibility studies, will be explained. The operational safety analysis is structured around physical components and real activities (construction, operation, closure) through a dedicated risk analysis. Due to the large variety of different ILW-LL waste, in order to identify the potential measures employed to counter each risk, there was a need to conduct the safety analysis in relation to four major hazards. These will be developed in the paper. The first one is the basic nuclear risk inherent in normal operations due to external exposure by irradiation (and also internal exposure by inhalation). Irradiation of ILW-LL waste packages are generally well above the 2 mSv/h level. The three other ones are related to the following particular accidental situations: the ILW-LL waste package drop during its emplacement in a disposal cell; the explosion associated with the emission of gas (due to hydrogen potential accumulation); and the fire breaking out in underground installations during construction or while in operation. The present paper will detail the above-mentioned items, the analysis methodology, the preliminary results of the risk analysis, and in particular the mitigating design measures taken. It will conclude with some suggestions regarding the ongoing development and design evolution that are needed in view of meeting the 2015 goal of a licensing permit application in accordance of the June 28, 2006 French Waste Act. (author)

[en] In 2001 SKB prepared an R and D programme /1/ for the KBS-3H, horizontal emplacement design of the repository for encapsulated spent nuclear fuel as an alternative repository design to KBS-3V, vertical emplacement. Both are based on the so-called KBS-3 method with multi-barrier systems but the orientation of the canister differs. It should also be noted that KBS-3V is the reference repository design for spent nuclear fuel in Sweden. The repository for spent fuel is planned to be constructed at a depth of about 500 metres in crystalline bedrock. The total amount of spent fuel within the Swedish programme if the reactors are operated 50-60 years will be in the order of 9000 tonnes or about 6000 canisters. The deposition capacity is planned to be in the order of 160 canisters per year. Start-up for emplacement of canisters with encapsulated spent nuclear fuel is now scheduled for 2020. The purpose of this paper is to present preliminary analysis on operational safety for a KBS-3H repository highlighting the two main areas where the KBS-3H and KBS-3V designs differ. The areas are the activities in the reloading station and in the disposal drift. (author)

[en] With almost a decade of operating experience, the Waste Isolation Pilot Plant (WIPP) has established an enviable record by clearly demonstrating that a deep geologic repository for unconditioned radioactive waste in rock salt can be operated safely and in compliance with very complex regulations. WIPP has disposed of contact-handled transuranic (TRU) waste since 1999 and remote-handled TRU waste since 2007. Emplacement methods range from directly stacking unshielded 0.21-4.5 m³ containers inside disposal rooms to remotely inserting highly radioactive 0.89 m³ canisters into horizontally drilled holes (shield plugs placed in front of canisters protect workers inside active disposal rooms). More than 100 000 waste containers have been emplaced, and one-third of WIPP's authorized repository capacity of 175,000 m³ has already been consumed. Principal surface operations are conducted in the waste handling building, which is divided into CH and RH waste handling areas. Four vertical shafts extend from the surface to the disposal horizon, 655 m below the surface in a 1000 m thick sequence of Permian bedded salt. The waste disposal area of about 0.5 km² is divided into ten panels, each consisting of seven rooms. Vertical closure (creep) rates in disposal rooms range up to 10 cm per year. While one panel is being filled with waste, the next one is being mined. Mined salt is raised to the surface in the salt shaft, and waste is lowered down the waste shaft. Both of these shafts also serve as principal access for personnel and materials. Underground ventilation is divided into separate flow paths, allowing simultaneous mining and disposal. A filter building near the exhaust shaft provides the capability to filter the exhaust air (in reduced ventilation mode) through HEPA filters before release to the atmosphere. WIPP operations have not exposed employees or the public to radiation doses beyond natural background variability. They consistently meet or exceed regulatory standards and expectations. Process improvements continuously reduce cycle times and costs. During the past few years, regulators have approved configuration changes that eliminated some unnecessary tests and activities. Many more could be targeted to further reduce vulnerability, while maintaining and even enhancing safety. While WIPP is licensed to dispose of only defense-related TRU waste, past experiments and performance assessments have shown that heat-generating high-activity waste could also be safely isolated in salt (and without prior vitrification). Thus, beyond its current restrictions, WIPP helps pave the way toward permanent isolation of all categories of radioactive waste. (author)

[en] Bentonite is a well known sealing and buffer material for waste disposals in hard rock conditions. Bentonite was also tested as a sealing material for rock salt formations. Under salt conditions the hydraulic conductivity is higher and the swelling pressure is lower then under freshwater conditions using the same bentonite dry density. Because rock salt conditions are more ambitious for bentonite, the new research results of material design and emplacement technology are very useful for applications under freshwater conditions. Commercial available bentonites were tested by measurements of hydraulic conductivity and swelling pressure at various dry densities. The result is the required emplacement dry density for each bentonite. Following an industrial production process for high compacted bentonite materials with low initial water content was developed. Bentonite blocks for drift sealings and compacts / granules for bulk mixtures for shaft sealings have been successful tested in large scale in situ tests in salt mines. Since 1998 about 500 t of bentonite blocks have been produced by the company Preiss-Daimler Industries GmbH - Feuerfestwerke Wetro. The blocks have a standard size of (250 x 125 x 62,5) mm. A proper fit of the blocks to the rock contour can be formed by sawing. The emplacement as dry brickwork is simple and reliable for the rough conditions in underground mines. The recent shaft sealing systems consist of a binary mixture of air dry compacted bentonite compacts and granules (moisture content 7 - 10 %). This material design and the production technology were developed in cooperation with the K and S Group. Both components of the bulk mixture (compacts and granules) are now produced at the plant 'Bergmannssegen-Hugo' (K and S Group). This material defines the actual best state of the art for bentonite sealing materials for long term stable shaft sealing systems, especially under difficult conditions like in salt mines. Since 2004 about 3000 t of bentonite compacts and high quality granules have been produced. For this reason 3 shaft sealing systems (Salzdetfurth mine) were realised. Furthermore 8 shaft sealings are intended to be build up to 2015 (closed mine 'Merkers'). With compacted granulate of various grain size a wide range of bentonite-sand-mixtures with defined hydraulic conductivity and swelling pressure may be designed. The wide experience with this bentonite materials (blocks, various bulk mixtures) is very useful for applications as buffer and sealing materials for radioactive waste disposals. (author)

[en] Both Sweden and Finland has advanced plans for design, construction and operation of the final repositories for direct disposal of spent nuclear fuel. Both countries have the same type of host rock - granite. They are also investigating alternative concept for disposal, vertical or horizontal disposal of the canisters with encapsulated spent nuclear fuel, normally called KBS-3V or the KBS-3H disposal concept. The development of the KBS-3V concept started around 1980 and is the reference method for both SKB in Sweden and Posiva in Finland. However, extensive development work is ongoing since 2001 with KBS-3H in order to bring that concept to the same maturity as KBS-3V. This presentation deals with the design and operation of the KBS-3V based on the work done within Sweden and SKB but the development is Finland is identical and it is a close cooperation between SKB in Sweden and Posiva in Finland. In Sweden, the site investigation for location of the repository has been concentrated on two sites, in the Oskarshamn area, about 350 km south of Stockholm, and the Forsmark area, about 180 km north of Stockholm. For information it can be mentioned that Finland plans to locate their repository in the vicinity of the Olkiluoto nuclear power plant site, about 300 km north of Helsinki. The site investigation is completed and the selection of site is scheduled to mid 2009 and sending in the application for location and construction of the repository is scheduled to end 2009. After receiving all necessary permits, construction time and commissioning will take about 7 to 8 years and operation is expected to start about 2020. The KBS-3 system is based on a multi barrier concept and the work with compiling the design requirements for the underground part of the deep repository has been ongoing for some time within the SKB organisation. Today the design requirements for the underground part are documented in a big number of reports that has been produced by specialists and working groups over rather a long time period. For each barrier the following will be determined during the development work: specification; design determining parameters; dependency on other parts of the repository barriers; design determining situations during construction and operation; and design determining processes after closure of the repository. For the design and optimisation of the different parts of the disposal concept the following has to be considered: long term safety after closure; safety during operation; safety during construction; environment issues; technology and feasibility; costs; possibility to retrieve the canisters with spent fuel. Over the years, a number of generic studies of the layout of the operational area(s) above ground and underground facilities of the repository have been performed. Different access routes from the ground level to the repository level at 500 m below ground have also been investigated. The access routes studied are mainly by shafts only or a ramp access for the heavy and bulky transports in combination with different service shafts. Further, a ramp alternative could be arranged as a spiral or as a straight ramp in combination with service shafts. The selected reference alternative with a combination of shafts and a spiral ramp is illustrated. In the evaluation process of the access routes it is important to have a good knowledge of the cost and time schedule for shaft sinking down to the repository level in order to compare this with a ramp access. The time needed for the excavation of the shaft down to the repository level will be about 18 months shorter comparing with ramp. Other factors that may influence the selection between access routes are constructability, operational safety and long term safety. In the reference case, the deposition tunnels are separated by 40 metres and the spacing between the deposition holes is six metres. The latter distance is determined by the need to limit the temperature on the canister surface. The operational area comprises a terminal building for receiving transport casks containing encapsulated fuel, a production building for preparation of buffer and backfill material, a supply building for electric power supply, buildings for offices and personnel and a restaurant building. When all canisters in one deposition drift have been emplaced the backfilling and final sealing of the drift can start. It has been discussed to carry out stepwise backfilling but the reference is that we do the backfilling in one step. The backfilling will be done with pre-compacted blocks of swelling clay and with some additional pellets for filling the void between the blocks and the rock wall and the roof of the drift. The principle for emplacement of the blocks is still not decided but different methods and equipment will be tested. The backfilling of the about 300 m long disposal drifts will be a challenge. The speed for backfilling must be high as we must avoid piping and problem with water. We plan to take down about 350 - 400 tons of backfilling material per 24 hours. The backfilling of one drift is estimated to take 10 - 12 full weeks working all days in the week and around the clock. The transport logistic for the backfill material from the production building in the operational area on ground down to the repository level and out into the drift and feeding to the emplacement equipment as well as filling the void between the blocks and the walls and the roof will not be an easy task. When the backfilling is completed it is time for construction of the sealing plug. SKB is investigating different designs of this plug. The plug will be a cast low-pH concrete plug but it is still open if it will be a short reinforced plug or a longer taped plug or if we need a bentonite plug between the backfill and the concrete construction. (author)

[en] PEM was investigated as a full-scale demonstration for the design, manufacturing and construction by using simulated buffer material and overpack in consideration of horizontal emplacement. Also near full-scale tests were conducted to examine the applicability of air-bearing system which can be used to transport a heavy load at the drift tunnel as for PEM. With regard to PEM casing, design requirements were selected from the viewpoints of EBS performance and operation safety issues. The construction procedure was examined in consideration of the shapes of buffer material, which are previously positioned inside the casing. And design procedure of the casing was also examined and presented. A full-scale PEM casing as a longitudinally two-part divided cylinder type with connection flanges was manufactured by using carbon steel plate. The wall thickness of this non-leak tight type PEM casing was evaluated its mechanical integrity by 2-dimensional stress analysis in consideration of the emplacement condition on the drift tunnel basement. Mechanical integrity of a percolated type casing was also examined its mechanical integrity. Air-bearing unit, which originally apply to a flat/smooth surface, was modified to fit a curved surface of the drift tunnel. Two units were aligned with two parallel lines, which estimate to be able to lift 12 tons, about two-fifth of the total weight of full scale PEM. On the conducted transportation tests of the air-bearing units, considering the surface roughness of the drift tunnel, especially for its unevenness, capability and availability of the run-over such gaps were investigated. And effect of covering sheets which can improve the gapped surface into relatively smooth was also examined by using several candidate materials. Through these tests, combination of the covering sheets and the maximum available height difference were evaluated and identified. Also the maximum traction force to toe the loading was measured to design the air-bearing system. (author)

[en] The publication contains the text of 26 oral presentations and 4 poster presentations, all of which have been input to INIS. (P.A.)