During the operation and dismantling of nuclear plants, a variety of radioactive liquid wastes accumulate, which must be supplied to a disposal chain suitable for final disposal. In the initial conditioning step, with regard to their chemical properties these waste waters are usually adjusted by water treatment systems (predominantly evaporation facilities) permanently installed in nuclear power plants, and then, they are concentrated up to a predefined content of solids. Afterwards, the developing evaporator concentrate is desiccated locally and suitable for final disposal by own permanently installed conditioning installations or by means of mobile plants. For a mobile conditioning of evaporator concentrates, GNS uses the drying plant of the type FAVORIT proven over decades. Each of the mobile FAVORIT processing plants is provided with a handling licence for Germany valid nationwide according to § 7 of the German Radiation Protection Ordinance (StrlSchV).

In this paper, we discuss the experiences during the construction of the first LILW disposal facility in South Korea. In December 2005, the South Korean Government designated Gyeongju-city as a host city of Low- and Intermediate-Level Radioactive Waste (LILW) disposal site through local referendums held in regions whose local governments had applied to host disposal facility in accordance with the site selection procedures. The LILW disposal facility is being constructed in Bongilri, Yangbuk-myeon, Gyeongju. The official name of the disposal facility is called ‘Wolsong Low and Intermediate Level Radioactive Waste Disposal Center (LILW Disposal Center)’. It can dispose of 800,000 drums of radioactive wastes in a site of 2,100,000 square meters. At the first stage, LILW repository of underground silo type with disposal capacity of 100,000 drums is under construction expected to be completed by June of 2014. The Wolsong Low and Intermediate Level Radioactive Waste Disposal Center consists of surface facilities and underground facilities. The surface facilities include a reception and inspection facility, an interim storage facility, a radioactive waste treatment building, and supporting facilities such as main control center, equipment & maintenance shop. The underground facilities consist of a construction tunnel for transport of construction equipment and materials, an operation tunnel for transport of radioactive waste, an entrance shaft for workers, and six silos for final disposal of radioactive waste. As of Dec. 2012, the overall project progress rate is 93.8%.

Large scale demonstration experiments in underground research laboratories (both onsite and off-site) are currently undertaken by most high level radioactive waste management organisations. The decision to plan and implement prototype experiments, which might have a life of several decades, has both important strategic and budgetary consequences for the organisation. Careful definition of experimental objectives based on the design and safety requirements is critical. The implementation requires the involvement of many parties and needs flexible but consequent management as, for example, additional goals for the experiments, identified in the course of the implementation, might jeopardise initial primary goals. The outcomes of an international workshop in which European and Japanese implementers (SKB, Posiva, Andra, ONDRAF, NUMO and Nagra) but also certain research organisations (JAEA, RWMC) participated identified which experiments are likely to be needed depending on the progress in implementing a disposal programme. Already earlier in a programme, large scale demonstrations are generally performed aiming at reducing uncertainties identified during the safety case development such as thermo-hydraulic-mechanical process validation in the engineered barrier system and target host rock. Also feasibility testing of underground construction in a potential host rock at relevant depth might be required. Later in a programme, i.e., closer to the license application, large scale experiments aim largely at demonstrating engineering feasibility and performance confirmation of complete repository components. Ultimately, before licensing repository operation, 1:1 scale commissioning testing will be required. Factors contributing to the successful completion of large scale demonstration experiments in terms of planning, defining the objectives, optimising results and main lessons learned over the last 30 years are being discussed. The need for international coordination in defining the objectives of new large scale demonstration experiments is addressed. The paper is expected to provide guidance to implementing organisations (especially those in their early stages of the programme), considering participating in and/or or conducting on their own large scale experiments in the near future.

In the frame of its radwaste disposal research programme, SCK•CEN started the construction of the HADES underground research facility in 1980. Including several extensions and a comprehensive experimental programme, it has provided a lot of experience on monitoring. Monitoring is performed for many reasons: construction follow-up, field characterisation, investigation of phenomena, and model validations — in which the underground lab offers the opportunity for upscaling conventional laboratory set-ups. Construction monitoring has allowed to develop and optimise the underground construction techniques in a previously poorly known environment, resulting in a well-mastered application of mechanised methods for gallery construction with minimal damage to the host formation. Access to this formation also allows its characterisation, both geotechnical, geological and geochemical, and the detailed investigation of phenomena such as fracturing and oxidation. Finally, instrumented set-ups allow to test various numerical models by comparing the observations with the predicted behaviour. The specific conditions of the underground laboratory put particular requirements to the sensors. These conditions include the long-term nature of many set-ups — typically several years to decades, the inaccessibility of many sensors after installation, high mechanical and water pressures, and corrosion. Combined with the fact that many sensors are custom made, obtaining and maintaining a fully functional instrumented set-up can be challenging. A lot of experience has therefore been gained which is very valuable when designing the monitoring of radwaste repositories — and it has allowed us to determine the critical success factors for monitoring. Engineers tend to look at this first from a technical viewpoint — and there are many technical aspects indeed that determine the reliability of monitoring. A first one is the combination of different observations (“redundancy”) which can be implemented by the use of several sensors, different sensor principles, different (coupled) parameters, and the combination of point measurements with geophysical techniques. Cabling is also a critical issue as it is often considered as the primary enemy of barrier integrity. Minimal cabling techniques, such as distributed fibre optic monitoring and wireless signal transmission, therefore get increasing attention. Also the interpretation of the monitoring data — in particular those that are perceived as “wrong” or “unexpected”, needs sufficient attention. The long-term experience has however also shown that the design of a monitoring programme must look beyond the technical part. In particular for long-term applications, issues such as data management and record keeping are vital to guarantee success in this.

ONDRAF/NIRAS, the Belgian Agency for Radioactive Waste and Enriched Fissile Materials, and its partners have developed long-term safety and performance analyses in the framework of the license application for a surface disposal facility for low level radioactive waste (category A waste) at Dessel, Belgium. This paper focusses on the methodology of the safety assessments and on key results from the application of this methodology. An overview is given (1) of the performance analyses for the containment safety function of the disposal system and (2) of the radiological impact analyses confirming that radiological impacts are below applicable reference values and constraints and leading to radiological criteria for the waste and the facility. In this discussion, multiple indicators for performance and safety are used to illustrate the multi-faceted nature of long-term performance and safety of the surface disposal. This contributes to the multiple lines of reasoning for confidence building that a positive decision to proceed to the next stage of construction is justified.

The paper describes the design methodology that ONDRAF/NIRAS, the Belgian Agency for Radioactive Waste and Enriched Fissile Materials, has developed and applied in the framework of the surface disposal facility for category A waste, i.e. solid conditioned low level radioactive waste that contains a limited amount of long lived radionuclides, at Dessel. The proposed disposal facility supporting the submitted construction and operation license application [1, 2] is also presented.

On 31 Januray 2013, ONDRAF/NIRAS, the Belgian Agency for Radioactive Waste and Enriched Fissile Materials, has introduced the construction and operation license application for the surface disposal facility for category A waste at Dessel. The objective of this paper is to present the development of this national programme. First the key milestones and their impact on the development of the project will be highlighted. This includes in particular: 1) the turning point in 1994, when ONDRAF/NIRAS converts its purely techno-scientific work method into an approach of participation that allows for societal aspects to be taken into account and 2) the decision of the government in 2006 in favour of the surface disposal of category A waste in Dessel. Then the paper presents the pillars of the long-term safety of a surface disposal and the principles for site selection and how they were applied to the surface disposal facility for category A waste at Dessel. The pillars of the long-term safety of a surface disposal are: (1) the passive confinement and isolation provided by the Engineering Barrier System, (2) the contribution to this passive isolation and confinement by the site, (3) the limitation of the radioactive source term, and (4) the control and monitoring measures.

The main aim of a gas-permeable seal is to increase the gas transport capacity of the backfilled underground structures without compromising the radionuclide retention capacity of the engineered barrier system or the host rock. Such a seal, proposed by NAGRA as part of the “Engineered Gas Transport System” in a L/ILW repository, considers specially designed backfill and sealing materials such as sand/bentonite (S/B) mixtures with a bentonite content of 20–30%. NAGRA’s RD&D plan foresees demonstrating the construction and performance of repository seals and improving the understanding and the database for reliably predicting water and gas transport through these systems. The fluid flow and gas transport properties of these backfills have been determined at the laboratory scale and through modelling the maximum gas pressures in the near field of a repository system and the gas flow rates have been evaluated. Within this context, the Gas-permeable Seal Test (GAST) was constructed at Grimsel Test Site (GTS) to validate the effective functioning of gas-permeable seals at realistic scale. The intrinsic permeability of such seals should be in the order of 10 −18 m 2 . Because the construction of S/B seals is not common practice for construction companies, a stepwise approach was followed to evaluate different construction and quality assurance methods. As a first step, an investigation campaign with simple tests in the laboratory and in the field followed by 1:1 scale pre-tests at GTS was performed. Through this gradual increase of the degree of complexity, practical experience was gained and confidence in the methods and procedures to be used was built, which allowed reliably producing and working with S/B mixtures at a realistic scale. During the whole pre-testing phase, a quality assurance (QA) programme for S/B mixtures was developed and different methods were assessed. They helped to evaluate and choose appropriate emplacement techniques and methodologies to achieve the target S/B dry density of 1.70 g/cm 3 , which results in the desired intrinsic permeability throughout the experiment. The final QA methodology was targeted at engineering measures to decide if the work can proceed, and at producing high resolution material properties database for future water and gas transport modelling activities. The different applied QA techniques included standard core cutter tests, the application of neutron-gamma (Troxler) probes and two mass balance methods (2D and 3D). The methods, looking at different representative scales, have provided only slightly different results and showed that the average density of the emplaced S/B plug was between 1.65 and 1.73 g/cm 3 . Spatial variability of dry densities was observed at decimeter scale. Overall, the pre-testing and QA programme performed for the GAST project demonstrated how the given design criteria and requirements can be met by appropriately planning and designing the material emplacement.

The first low- and intermediate-level waste (LILW) disposal facility is under construction in saturated granite in Korea. The safety assessment report (SAR) identified that different gases, such as hydrogen, carbon dioxide, and methane are generated at the disposal facility due to the corrosion of metal wastes and steel drum, and microbial degradation of organic matters. Reinforced concrete plays a role as an engineered barrier at the disposal facility, so its properties with regard to gas migration were evaluated in laboratory-scale experiments. Then modeling of gas migration was carried out to evaluate gas pressure build-up in the disposal facility. The gas entry pressure and relative gas permeability of the concrete was determined to be 0.97±0.15 bar, and the relative gas permeability decreased exponentially with increasing water content. The results of the modeling showed that most of hydrogen gas was dissolved in groundwater and did not significantly influence pressure build-up inside the disposal facility based on the reference case of gas generation.

In 2009, remediation was initiated for a non-operational fuel cycle facility previously used for government contract work located in Windsor, Connecticut, USA. Radiological contaminants consisted primarily of high enriched uranium (HEU). Other radionuclides encountered in relatively minor amounts in certain areas of the clean-up included Co-60, Cs-137, Ra-226, Th-232 and low enriched uranium (LEU). Between 2009 and the spring of 2011, remediation efforts were focused on demolition of contaminated buildings and removal of contaminated soil. In the late spring of 2011, the last phase of remediation commenced involving the removal of contaminated sediments from portions of a 1,200 meter long gaining stream. Planning and preparation for remediation of the stream began in 2009 with submittal of permit applications to undertake construction activities in a wetland area. The permitting process was lengthy and involved securing permits from multiple agencies. However, early and frequent communication with stakeholders played an integral role in efficiently obtaining the permit approvals. Frequent communication with stakeholders throughout the planning and remediation process also proved to be a key factor in timely completion of the project. The remediation of the stream involved the use of temporary bladder berms to divert surface water flow, water diversion piping, a sediment vacuum removal system, excavation of sediments using small front-end loaders, sediment dewatering, and waste packaging, transportation and disposal. Many safeguards were employed to protect several species of concern in the work area, water management during project activities, challenges encountered during the project, methods of Final Status Survey, and stream restoration.

We are having 23 units of nuclear power plants in operation and 5 units of nuclear power plants under construction in Korea as of September 2012. However, we don’t have any experience on shutdown permanently and decommissioning of nuclear power plants. There are only two research reactors being decommissioned since 1997. It is realized that improvement of the regulatory framework for decommissioning of nuclear facilities has been emphasized constantly from the point of view of IAEA’s safety standards. It is also known that IAEA will prepare the safety requirement on decommissioning of facilities; its title is the Safe Decommissioning of Facilities, General Safety Requirement Part 6. According to the result of IAEA’s Integrated Regulatory Review Service (IRRS) mission to Korea in 2011, it was recommended that the regulatory framework should require decommissioning plans for nuclear installations to be constructed and operated and these plans should be updated periodically. In addition, after the Fukushima nuclear disaster in Japan in March of 2011, preparedness for early decommissioning caused by an unexpected severe accident became important issues and concerns. In this respect, it is acknowledged that the regulatory framework for decommissioning of nuclear facilities in Korea need to be improved. First of all, we focus on identifying the current status and relevant issues of regulatory framework for decommissioning of nuclear power plants compared to the IAEA’s safety standards in order to achieve our goal. And then the plan is established for improvement of regulatory framework for decommissioning of nuclear power plants in Korea. It is expected that if the things will go forward as planned, the revised regulatory framework for decommissioning could enhance the safety regime on the decommissioning of nuclear power plants in Korea in light of international standards.

On 26 April 1986, the worst nuclear power plant accident in history occurred at the Chernobyl plant in Ukraine (then part of the Soviet Union). The destruction of Unit 4 sent highly radioactive fallout over Belarus, Russia, Ukraine, and Europe. The object shelter — a containment sarcophagus — was built in November 1986 to limit exposure to radiation. However, it has only a planned 25-year lifespan and would probably not survive even a moderate seismic event in a region that has more than its share of such events. It was time to take action. One of the largest tasks that are in progress is the design and construction of the New Safe Confinement (NSC). The NSC is an engineered enclosure for the entire object shelter that includes a suite of process equipment. The process equipment will be used for the dismantling of the destroyed Chernobyl Nuclear Power Plant Unit. One of the major mechanical handling systems to be installed in the new safe confinement is the Main Cranes System. The planned decontamination and decommissioning or dismantling activities will require the handling of heavily shielded waste disposal casks containing nuclear fuel as well as lifting and transporting extremely large structural elements. These activities, to be performed within the new safe confinement, will require large and sophisticated cranes. The article will focus on the current progress of the new safe confinement and of the main cranes system for the decommissioning or dismantling activities.

The underwater spectrometric system for survey the bottom of material science multi-loop reactor MR ponds was elaborated. This system uses CdZnTe (CZT) detectors that allow for spectrometric measurements in high radiation fields. The underwater system was used in the spectrometric survey of the bottom of the MR reactor pool, as well as in the survey located in the MR storage pool of highly radioactive containers and parts of the reactor construction. As a result of these works irradiated nuclear fuel was detected on the bottom of pools, and obtained estimates of the effective surface activity detected radionuclides and created by them the dose rate.

The research reactor Thetis, in service since 1967 and stopped in 2003, is part of the laboratories of the institution of nuclear science of the University of Ghent. The reactor, of the pool-type, was used as a neutron-source for the production of radio-isotopes and for activation-analyses. The reactor is situated in a water pool with inner diameter of 3 m. and a depth of 7.5 m. The reactor core is situated 5.3m under water level. Besides the reactor, the pool contains pneumatic loops, handling tools, graphite blocks for neutron moderation and other experimental equipment. The building houses storage rooms for fissile material and sources, a pneumatic circuit for transportation of samples, primary and secondary cooling circuits, water cleaning resin circuits, a ventilation system and other necessary devices. Because of the experimental character of the reactor, laboratories with glove boxes and other tools were needed and are included in the dismantling program. The building is in 3 levels with a crawl-space. The ground-floor contains the ventilation installation, the purification circuits with tanks, cooling circuits and pneumatic transport system. On the first floor, around the reactor hall, the control-room, visiting area, end-station for pneumatic transport, waste-storage room, fuel storage-room and the labs are located. The second floor contains a few laboratories and end stations of the two high speed transfer tubes. The lowest level of the pool is situated under ground level. The reactor has been operated at a power of 150 kW and had a max operating power of 250 kW. Belgoprocess has been selected to decommission the reactor, the labs, storage halls and associated circuits to free release the building for conventional reuse and for the removal of all its internals as legal defined. Besides the dose-rate risk and contamination risk, there is also an asbestos risk of contamination. During construction of the installation, asbestos-containing materials were used, which must be removed in controlled conditions. The ventilation system is considered free from nuclear contamination but it contains asbestos. This paper covers the organization of the dismantling work, the technical execution aspect and conclusions already known (dismantling is ongoing as this is written).

A high temperature gas cooled reactor-pebble bed module (HTR-PM) is currently under design and construction in China. Calculations of induced radioactivity in the concrete containment vessel of HTR-PM were performed in order to predict the decommissioning problems. The main activation products and the ambient dose equivalent rate corresponding to 40 years of operations followed by 30 days/1 year/10 years of decay, were calculated by using the Monte Carlo particle transport code FLUKA. Finally, according to the Chinese National Standard and on the basis of the knowledge gained from this study, the classifications of the radioactive solid wastes were discussed, and some suggestions for the working duration of the employees were given.

The Jose Cabrera Nuclear Power Plant (NPP) was the first commercial power reactor (Westinghouse 1 loop PWR 510 MWth, 160 MWe) commissioned in Spain and provided the base for future development and training. The reactor construction started in 1963 and it was officially on-line by 1969. The NPP operated from 1969 until 2006 when it became the first reactor to be shut down after completing its operational period. The containment is reinforced concrete with a stainless steel head. In 2010 responsibility for D&D was transferred to Enresa to achieve IAEA level 3 (a green field site available for unrestricted re-uses) by 2017. Of the total of more than 104,000 tons of materials that will be generated during dismantling, it is estimated that only ∼ 4,000 tons will be radioactive waste, some of which, 40 t are considered as intermediate level long-lived wastes and the rest (3,960 t) will be categorized as VLLW & ILLW. The Project is divided into five phases: Phase 0 - Removal of fuel and preliminary work. Phase 1 - Preparatory Activities for D&D. complete. Phase 2 - Dismantling of Major Components. Phase 3 - Removal of Auxiliary Installations, Decontamination and Demolition. Phase 4 - Environmental Restoration. Phase 2, is currently ongoing (50% completed). To manage the diverse aspects of decommissioning operations, Enresa uses an internally developed computerized project management tool. The tool, based on knowledge gathered from other Enresa projects, can process operations management, maintenance operations, materials, waste, storage areas, procedures, work permits, operator dose management and records. Enresa considers that communication is important for both internal and external stakeholder relations and can be used to inform, to neutralize negative opinions & attitudes, to remove false expectations and for training. Enresa has created a new multi-purpose area (exhibition/visitor centre) and encourages visits from the public, local schools, local and national politicians and technical groups. Greenfield is the final end state objective. The total cost of this project, including a 20% contingency as estimated in 2003 is 135 M€. This figure does not include the management of the plant spent fuel, which has constituted an independent project that has been completed in 2009 (35 M€). Enresa, with 15 staff on site are managing a team of ∼ 250 workers, 40 of whom belong to the previous operator. The spent fuel is On-Site prior to the final destination in the future Spain Centralized Spent Fuel Installation.

Ningyo-toge Uranium Mine is subject to the environmental remediation. The main purposes are to take measures to ensure the radiation protection from the exposure pathways to humans in future, and to prevent the occurrence of mining pollution. The Yotsugi Mill Tailings Pond in the Ningyo-toge Uranium Mine has deposited mining waste and impounded water as a buffer reservoir before it is transferred to the Water Treatment Facility. It is located at the upstream of the water-source river and as the impact on its environment in case of earthquake is estimated significant, the highest priority has been put to it among mine-related facilities in the Mine. So far, basic concept has been examined and a great number of data has been acquired, and using the data, some remediation activities have already done, including capping construction for the upstream part of the Mill Tailings Pond. The capping is to reduce rainwater penetration to lower the burden of water treatment, and to reduce radon exhalation and dose rates. Only natural materials are used to alleviate the future maintenance. Data, including settlement amount and underground temperature is now being acquired and accumulated to verify the effectiveness of the capping, and used for the future remediation of the Downstream with revision of its specifications if necessary.

Geological hazard assessments are being used to make important decisions relevant to nuclear facilities such as a repository for deep geological disposal of high-level radioactive waste. With respect to such repositories, topographic evolution is a key issue for description of the long-term evolution of a groundwater flow characteristics in time spans of tens to hundreds of thousands of years. The construction of topographic evolution models is complex, involving tacit knowledge and working processes. Therefore, it is important to externalise, that is to explicitly present the tacit knowledge and decision-making processes used by experts in the model building unambiguously, with thorough documentation and to provide key knowledge to support planning and implementation of investigations. In this study, documentation of the technical know-how used for the construction of a topographic evolution model is demonstrated. The process followed in the construction of the model is illustrated using task-flow logic diagrams; the process involves four main tasks with several subtasks. The task-flow followed for an investigation to estimate uplift rates linked to the task-flow for the modelling of topographic evolution is also illustrated. In addition, the decision-making processes in the investigation are expressed in logical IF-THEN format for each task. Based on the documented technical know-how, an IT-based Expert System was constructed. In future work, it is necessary to analyse the knowledge, including the management of uncertainties in the modelling and investigations, and to integrate fundamental ideas for managing uncertainties with expert system.

In Germany low / medium level waste, which is classified here as radioactive waste with negligible heat generation, will be disposed of in the Konrad underground repository. The construction and the operation of this nuclear facility required authorization by different fields of law, i.e., by nuclear law, mining law and water law. Whereas the nuclear law considers solely radiological aspects, the relevant permit issued according to the water law considers the impact of radioactive as well as non-radioactive harmful substances. The Federal Office for Radiation Protection (BfS) as operator of the repository and permit holder has (a) to record the disposed of radioactive and non-radioactive harmful substances and (b) to balance them. To meet these requirements BfS has developed a concept, which led to a site specific solution. Threshold values were defined for recording and for balancing the harmful substances. It had to be verified that by disposal of radioactive waste packages according to these values an adverse effect on the near-surface groundwater can be excluded. The Lower Saxony Water Management, Coastal Protection and Nature Conservation Agency (NLWKN) as the responsible water law regulatory authority approved the operator’s concept as appropriate to comply with the requirements of the Water Law Permit. Nonetheless, collateral clauses were imposed to assure this.

Svensk Ka¨rnbra¨nslehantering AB (SKB) has performed comprehensive investigations of two candidate sites for a final repository for Sweden’s spent nuclear fuel. In March 2011 SKB decided to submit licence applications for a final repository at Forsmark. Before selection, SKB stated that the site that offers the best prospects for achieving long-term safety in practice would be selected. Based on experiences previous safety assessments, a number of issues related to long-term safety need to be considered in the context of site comparison. The factors include sensitivity to climate change such as periods of permafrost and glaciations, rock mechanics evolution including the potential for thermally induced spalling and sensitivity to potential future earthquakes, current and future groundwater flow, evolution of groundwater composition and proximity to mineral resources. Each of these factors related to long-term safety for the two candidate sites is assessed in a comparative analysis of site characteristics. The assessment also considers differences in biosphere conditions and in the confidence of the site descriptions. The comparison is concluded by an assessment on how the identified differences would affect the estimated radiological risk from a repository located at either of the sites. The assessment concludes that there are a number of safety related site characteristics for which the analyses do not show any decisive differences in terms of implications on safety, between the sites Forsmark and Laxemar. However, the frequency of water conducting fractures at repository depth is much smaller at Forsmark than at Laxemar. This difference, in turn, affects the future stability of the current favourable groundwater composition, which combined with the much higher flows at Laxemar would, for the current repository design, lead to a breach in the safety functions for the buffer and the canister for many more deposition positions at Laxemar than at Forsmark. Thereby the calculated risk for Forsmark will be considerably lower than that for Laxemar. What decided the choice is that Forsmark is thus mainly that it was judged to offer better prospects for achieving long-term safety in the final repository. Other factors considered included implications regarding repository construction and operation, difference in the footprint of the repository, comparisons regarding environment and health as well as social resources and local support.