The Supercontainer (SC) design is the preferred option for the final disposal of high level nuclear waste and spent fuel in Belgium. The SC consists of a carbon steel overpack surrounded by a very thick concrete buffer, contained within a stainless steel envelope. In this highly alkaline environment of the annulus and under normal conditions (without the ingress of aggressive species), the carbon steel overpack will be protected by a passive oxide film, which is believed to result in very low and almost negligible uniform corrosion rates. This paper discusses the modeling efforts simulating the evolution of various parameters that can potentially influence the corrosion processes (temperature, pH, saturation degree and concentration of aggressive species). The outcome of the Belgian experimental program to study the passive uniform corrosion of carbon steel in concrete is also addressed. Long-term uniform corrosion rates are determined using different independent techniques (hydrogen gas measurements by means of manometric gas cells or pressure transducers, long-term passive current density measurements).

The paper aims to highlight recent developments at the Belgian Nuclear Research Centre SCK•CEN in experimental and numerical study of the coupled physical-chemical behaviour of concrete subject to chemical degradation. The discussion mainly focusses on three interlinked research projects covering novel experimental methods to study the alteration of hydraulic and transport properties during carbonation and calcium leaching, a pore scale numerical model to capture microstructural changes due to the above degradation processes and a generic multiscale model aimed at determining evolution of the properties of a macrostructure over the long term. The paper also describes supplementary continuum scale numerical studies concerning concrete-clay interactions and geochemical impact on the physical structure of concrete. Preliminary findings from these studies show encouraging results such as the development of novel leaching, water permeability and diffusion apparatus, a robust pore scale model based on Lattice-Boltzmann method and a mesoscale study focused on the importance of interfacial transition zones on the effective diffusivity for linear and nonlinear diffusion problems.

The aircraft impact accident has become very significant in the design of a nuclear facilities, particularly, after the tragic September 2001 event, that raised the public concern about the potential damaging effects that the impact of a large civilian airplane could bring in safety relevant structures. The aim of this study is therefore to preliminarily evaluate the global response and the structural effects induced by the impact of a military or commercial airplane (actually considered as a “beyond design basis” event) into a near surface radioactive waste (RWs) disposal facility. The safety evaluation was carried out according to the International safety and design guidelines and in agreement with the stress tests requirements for the security track. To achieve the purpose, a lay out and a scheme of a possible near surface repository, like for example those of the El Cabril one, were taken into account. In order to preliminarily perform a reliable analysis of such a large-scale structure and to determine the structural effects induced by such a types of impulsive loads, a realistic, but still operable, numerical model with suitable materials characteristics was implemented by means of FEM codes. In the carried out structural analyses, the RWs repository was considered a “robust” target, due to its thicker walls and main constitutive materials (steel and reinforced concrete). In addition to adequately represent the dynamic response of repository under crashing, relevant physical phenomena (i.e. penetration, spalling, etc.) were simulated and analysed. The preliminary assessment of the effects induced by the dynamic/impulsive loads allowed generally to verify the residual strength capability of the repository considered. The obtained preliminary results highlighted a remarkable potential to withstand the impact of military/large commercial aircrafts, even in presence of ongoing concrete progressive failure (some penetration and spalling of the concrete wall) of the impacted area.

The study of embedding matrix of radioactive waste is an important issue concerning the need to assure the radiological safety requirements for final disposal. The conditioning of the radioactive wastes by cementation process involves also a structural investigation by X-ray diffraction (XRD) of the samples prepared with cement and complexing agents at different concentration. This paper gives useful information about the influence of complexing agents related to damages produced in concrete microstructures. The experimental data obtained on samples kept in real repository environment for a period up to 15 years, offer structural information regarding the interaction between the different decontamination agents with the major components of the matrix.

Since the 1950’s radioactive wastes from a number of laboratories have been stored below ground at the Hanford site, Washington State, USA, in vertical pipe units (VPUs) made of five 200 litre drums without tops or bottoms, and in caissons, made out of corrugated pipe, or concrete and typically 2,500mm in diameter. The VPU’s are buried of the order of 2,100mm below grade, and the caissons are buried of the order of 6,000mm below grade. The waste contains fuel pieces, fission products, and a range of chemicals used in the laboratory processes. This can include various energetic reactants such as un-reacted sodium potassium (NaK), potassium superoxide (KO 2 ), and picric acid, as well as quantities of other liquids. The integrity of the containments is considered to present unacceptable risks from leakage of radioactivity to the environment. This paper describes the successful development and full scale testing of in-ground augering equipment, grouting systems and removal equipment for remediation and removal of the VPUs, and the initial development work to test the utilization of the same basic augering and grouting techniques for the stabilization, size reduction and removal of caissons.

Radioactive waste systems and structures (RWSS) are safety-critical facilities in need of monitoring over prolonged periods of time. Structural health monitoring (SHM) is an emerging technology that aims at monitoring the state of a structure through the use of networks of permanently mounted sensors. SHM technologies have been developed primarily within the aerospace and civil engineering communities. This paper addresses the issue of transitioning the SHM concept to the monitoring of RWSS and evaluates the opportunities and challenges associated with this process. Guided wave SHM technologies utilizing structurally-mounted piezoelectric wafer active sensors (PWAS) have a wide range of applications based on both propagating-wave and standing-wave methodologies. Hence, opportunities exist for transitioning these SHM technologies into RWSS monitoring. However, there exist certain special operational conditions specific to RWSS such as: radiation field, caustic environments, marine environments, and chemical, mechanical and thermal stressors. In order to address the high discharge of used nuclear fuel (UNF) and the limited space in the storage pools the U.S. the Department of Energy (DOE) has adopted a “Strategy for the Management and Disposal of Used Nuclear Fuel and High-Level Radioactive Waste” (January 2013). This strategy endorses the key principles that underpin the Blue Ribbon Commission’s on America’s Nuclear Future recommendations to develop a sustainable program for deploying an integrated system capable of transporting, storing, and disposing of UNF and high-level radioactive waste from civilian nuclear power generation, defense, national security, and other activities. This will require research to develop monitoring, diagnosis, and prognosis tools that can aid to establish a strong technical basis for extended storage and transportation of UNF. Monitoring of such structures is critical for assuring the safety and security of the nation’s spent nuclear fuel until a national policy for closure of the nuclear fuel cycle is defined and implemented. In addition, such tools can provide invaluable and timely information for verification of the predicted mechanical performance of RWSS (e.g. concrete or steel barriers) during off-normal occurrence and accident events such as the tsunami and earthquake event that affected Fukushima Daiichi nuclear power plant. The ability to verify the conditions, health, and degradation behavior of RWSS over time by applying nondestructive testing (NDT) as well as development of nondestructive evaluation (NDE) tools for new degradation processes will become challenging. The paper discusses some of the challenges associated to verification and diagnosis for RWSS and identifies SHM technologies which are more readily available for transitioning into RWSS applications. Fundamental research objectives that should be considered for the transition of SHM technologies (e.g., radiation hardened piezoelectric materials) for RWSS applications are discussed. The paper ends with summary, conclusions, and suggestions for further work.

Up to the end of this decade, corresponding to the planned date of starting operation in the final disposal KONRAD for non-heat generating waste in Germany, a lot of efforts are needed to condition and package the radioactive waste in containers certified by BfS (Federal Institute for Radiation Protection). This waste is produced by public sector and industry as well as nuclear energy facilities, which result in more than a half of the actually declared quantity growing especially after the phase out decision of nuclear power production in Germany and the subsequent decommissioning of nuclear power plants. BAM (Federal Institute for Materials Research and Testing) acts as responsible authority on behalf of BfS for design testing under consideration of the KONRAD requirements. Within the assessment procedure of containers, BAM has to verify the application documents, including material qualification, container geometry, corrosion protection, leakage rate and operational and accidental loading, and to evaluate quality assurance measures. Besides the previous completed approvals for various containers such as steel sheet and cast iron box-shaped containers and concrete cylindrical ones, BAM is actually carrying assessments for other types such as cast iron cylindrical containers and “old” steel sheet box shaped ones. The so called “old” containers present already loaded containers without any KONRAD certification, currently stored at licensed interim storages. In the benefit of the container assessment, BAM operates design testing facilities for drop and fire tests which are also used for research objectives to improve and expand the evaluation methods such as research project ConDrop. The latter deals with numerical simulations and analyses tools for further precise predictions about unfavorable drop test scenarios, safety margins, and design sensitivities for steel sheet containers. Furthermore, during the assessment procedure, several specifications about the containers, the inventory to be disposed and the required safety level have been intensely debated by authorities and applicants. Based on its experience in qualifying containers, BAM has been commissioned to identify insufficiently specified aspects in the waste acceptance criteria and to propose clearer definitions with regard to the secondary regulations in the planning approval notice. This paper focuses on aspects of the contribution of BAM in certifying KONRAD waste containers. This will include the current state of design testing procedures, evaluation methods that are available or are being developed and the main topics within the KONRAD requirements being actually specified by BAM.

This paper presents a consistent approach for the development of a comprehensive data base of time-dependent hydraulic and transport parameters for concrete engineered barriers of the future Dessel near surface repository for low level waste. The parameter derivation is based on integration of selected data obtained through an extensive literature review, data from experimental studies on cementitious materials specific for the Dessel repository and numerical modelling using physically-based models of water and mass transport. Best estimate parameter values for assessment calculations are derived, together with source and expert range and their probability density function wherever the data was sufficient. We further discuss a numerical method for upscaling laboratory derived parameter values to the repository scale; the resulting large-scale effective parameters are commensurate with numerical grids used in models for radionuclide migration. To accommodate different levels of conservatism in the various assessment calculations defined by ONDRAF/NIRAS, several sets of parameter values have been derived based on assumptions that introduce different degrees of conservatism. For pertinent parameters, the time evolution of such properties due to the long-term concrete degradation is also addressed. The implementation of the consistent approach is demonstrated by considering the pore water diffusion coefficient as an example.

The Philippine Nuclear Research Institute (PNRI) in collaboration with the interagency technical committee on radioactive waste has been undertaking a national project to find a final solution to the country’s low to intermediate level radioactive waste. The strategy adopted was to co-locate 2 disposal concepts that will address the types of radioactive waste generated from the use of radioactive materials. This strategy is expected to compensate for the small volumes of waste generated in the Philippines as compared to countries with big nuclear energy programs. It will also take advantage of the benefits of a shared infrastructure and R&D work that accompany such project. The preferred site selected from previous site selection and investigations is underlain by highly fractured “andesitic volcaniclastics” mantled by residual clayey soil which act as the aquifer or water bearing layer. Results of investigation show that the groundwater in the area is relatively dilute and acidic. Springs at the lower elevations of the footprint also indicate acidic waters. The relatively acidic water is attributed to the formation of sulfuric acid by the oxidation of the pyrite in the andesite. A preliminary post closure safety assessment was carried out using the GMS MODFLOW and HYDRUS softwares purchased through the International Atomic Energy Agency (IAEA) technical assistance. Results from MODFLOW modeling show that the radionuclide transport follows the natural gradient from the top of the hill down to the natural discharge zones. The vault dispersion model shows a circular direction from the vaults towards the faults and eventually to the creeks. The contaminant transport from borehole shows at least one confined plume from the borehole towards the creek designated as Repo1 and eventually follows downstream. The influx of surface water and rainfall to the disposal vault was modeled using the HYDRUS software. The pressure head and water content at the base of the foundation layer and the bottom of the concrete is where a significant reduction in water content can be observed. It is also noted that water content and pressure remain constant after one year.

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.

The economic, environmental, social and societal constraints have pushed the nuclear industry to develop new dismantling and decommissioning (D&D) techniques in order to meet the current industrial requirements. Since 5 years, AREVA, in partnership with Air Liquide, has been working on a new cryogenic solution that can be used to achieve the main important D&D operations (i.e. cutting, surface decontamination and concrete scabbling). This solution, called NiThrow™, is based on the Nitrocision LLC concept and consists in spraying highly pressurized liquid nitrogen (3500 bar) at a very low temperature (−140°C) onto a surface. The main advantage in using liquid nitrogen instead of water is due to its rapid conversion into gaseous nitrogen that will avoid the generation of liquid effluent. The preliminary trials made in France at Saint Ouen l’Aumône and at the AREVA SICN facility by using NiThrow™ appeared to be very interesting and promising in terms of effectiveness, reduction of the dosimetry (i.e. ALARA principle) and adaptability to the different types of surfaces and materials to be treated. On an other hand, they also highlighted a couple of defects like a lack of reliability that is not compatible with a safety use in a nuclear environment. In 2009, with the aim of fixing all these issues, AREVA decided to start an extensive R&D program. This work has been essentially focused on the lancing tool, the insulation and the flexibility of the liquid nitrogen distribution network, but also on the design and the manufacturing of a waste collection unit and a dedicated carrier. Today, these R&D efforts have been totally completed. and resulted to the publication of 3 patents. Also, the collaboration, with Air Liquide and the American company Conco allows AREVA to propose to its customers a safe and reliable solution for their cleaning operations in France and all over the world.

SCK•CEN is the Belgian Nuclear Research Centre. Founded in the mid-fifties, it has accumulated experience and know-how in all fields of the nuclear power production: in the neutronics calculation, radiation protection, waste management, fuel performance and analysis, nuclear measurements, radiochemistry, reactor technology, etc. Since 1989, SCK•CEN has launched Decommissioning activities to deal with the Technical Liabilities created by 40 years of operation. The main projects started were: • the dismantling of the BR3 PWR reactor; • the dismantling of active laboratories and the decontamination of buildings for unrestricted reuse; • the management of the waste arising from the refurbishment activities of the BR2, especially the management of the high active beryllium matrix. In 1989, the BR3 reactor, a Pressurized Water Reactor, was selected by the European Commission as one of the four pilot dismantling projects in the framework of the EC five year RTD program on dismantling nuclear installations. Through this project, SCK•CEN has built up a broad know-how in dismantling and decommissioning operations. This know how concerns the decontamination for dose rate reduction and/or free release of materials, teleoperated techniques for cutting highly activated components of a reactor, concrete decontamination techniques, characterization techniques of radioactive waste or for free release of components and development of decommissioning management and recordkeeping of material streams and of nuclear material accountancy. SCK•CEN is now actively involved in other decommissioning projects in Belgium and in expertise abroad. After giving an overview of the main achievements and the perspectives of the decommissioning of the BR3 reactor, the paper intends to present the involvement of SCK•CEN in the other decommissioning projects and to give an overview of our activities and capacities.

Belgoprocess started the industrial decommissioning of the main process building of the former Eurochemic reprocessing plant in 1987. In 2008 the eastern part was torn down and in 2010 the central part of the building was removed. The demolition of the third and last part is planned in 2014. At that point the area will be restored to brown field conditions. During the 25 years of decommissioning the Eurochemic installations a lot of effort has been invested in improving the efficiency. The Belgian decommissioning policy is based on maximizing effort for decontamination with the goal of unconditional release and minimization of nuclear waste. As a result a number of installations have been implemented that reduce the processing costs, increase the output of decontamination and increase the speed of release measurements. • The abrasive blasting installation is used to remove a thin layer from metal components and concrete objects. • The concrete crushing and sampling installation is used as the final measurement before unconditional release of concrete debris. • The concrete spec installation is used to separate the contaminated part from the non contaminated part of concrete debris. As the end of the decommissioning project is approaching the time is right to look back and evaluate the course of the project. Looking back what were the key decisions during the project, what were important milestones and what could or should have been done differently knowing what we know now.

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.

A number of proven technologies applicable to the chemical and physical decontamination of radioactive and non-radioactive contaminants within the environmental remediation and radiological waste management sectors exist. Previous work generally acknowledges that these methods have limitations such as production of large volumes of waste, destruction of the substrate, complex safety considerations [1a] and application of special precautions to meet disposal acceptance criteria [2] . A method that removes a variety of contaminants from the surface and subsurface of porous materials, with minimal contaminated waste arisings, is highly desirable. TechXtract ® is a patented, sequential chemical extraction process developed to remove radionuclides, PCBs, and other hazardous organic and inorganic substances from solid materials such as concrete, brick, steel, and exotic metals [3] . The technology uses multifarious task-specific chemical formulations and engineered applications to achieve surface penetration and removal of the contaminants from the atomic voids of metals and other substrates, or the capillaries and gel pores of concretes. TechXtract ® is proven to remove a variety of contaminants from various substrates, allowing free release of the substrate as waste for disposal, or re-use, whilst producing minimal secondary waste. Data from testing of TechXtract’s capabilities and evidencing the technology’s efficacy during site based applied research and development is presented here.

In the frame of the decommissioning of nuclear power plants or laboratories, the penetration depth of the contamination in concrete surfaces: walls, floor, is unknown. Its knowledge requires sample analysis, that is time consuming and expensive. The main goal of the work is to define and evaluate a non-destructive measurement technique for the evaluation of the contamination depth in concrete. Estimation of accuracy of measurements for different contamination levels, time of measurement, value of natural radionuclides (NRN) concentration in concrete and background radiation dose were carried out. The type of relevant detector depends of selected limiting sensitivity, the weight and sizes of the device.

The paper relates to the determination of the amount of nuclear material contained in a closed, concrete lined vault at the Aktau fast breeder reactor in Kazakhstan. This material had been disposed into the vault after examination in an experimental hot cell directly above the vault. In order to comply with IAEA Safeguards requirements it was necessary to determine the total quantities of nuclear materials — enriched uranium and plutonium — that were held with Kazakhstan. Although it was possible to determine the inventory of all of the accessible nuclear material — the quantity remaining in the vault was unknown. As part of the Global Threat Reduction Programme the UK Government funded a project to determine the inventory of these nuclear materials in this vault. This involved drilling three penetrations through the concrete lined roof of the vault; this enabled the placement of lights and a camera into the vault through two penetrations; while the third penetration enabled a lightweight manipulator arm to be introduced into the vault. This was used to provide a detailed 3D mapping of the dose rate within the vault and it also enabled the collection of samples for radionuclide analysis. The deconvolution of the 3D dose rate profile within the vault enabled the determination of the gamma emitting source distribution on the floor and walls of the vault. The samples were analysed to determine the fingerprint of those radionuclides producing the gamma dose — namely 137 Cs and 60 Co — to the nuclear materials. The combination of the dose rate source terms on the surfaces of the vault and the fingerprint then enabled the quantities of nuclear materials to be determined. The project was a major success and enabled the Kazakhstan Government to comply with IAEA Safeguards requirements. It also enabled the UK DECC Ministry to develop a technology of national (and international) use. Finally the technology was well received by IAEA Safeguards as an acceptable methodology for future studies.

SOGETER is a waste conditioning process for Low Level radwaste (LLW) or Intermediate Level radwaste (ILW) such as sludge, soil, ashes, evaporator concentrate, concrete rubble, asbestos or sand. Such radioactive waste is usually solidified into a cement matrix, resulting in a 2 to 5 times volumetric waste increase, i.e. 1 m 3 of initial raw waste generates 2 to 5 m 3 of solidified waste. Sogeter melts the waste at high temperature - up to 2000 K, and produces a synthetic rock. The main component of the matrix is the waste itself; therefore 1 m 3 of initial raw waste generates only 0.2 to 0.5 m 3 of solidified waste. Compared to cementation, synthetic rock decreases the volume to be disposed of by a factor of 4 up to a factor of 25. By mixing different types of waste, or using additives, the composition of the waste is adjusted so that a fluid melt is obtained at temperatures less than 2000 K and the final 200 L ingot can be cooled within 2 days, without shattering or disaggregating. A wide range of compositions have been tested demonstrating that almost every type of waste can be conditioned with Sogeter. An industrial facility has been designed based on very robust and proven heating and off-gas treatment technology. Industrial tests have been carried out on more than 2 tons of simulated, non-radioactive waste, producing blocks of treated matter weighing up to 250 kg. During these tests all process parameters, including electrical consumption, throughput and robustness, have been checked.

The decommissioning process is a quite long and complicated stage who may take few years or decades to be achieved. Generally, this process involves the implementation of a large number of technologies dedicated to cutting and decontamination operations. Based on this finding, the Clean-Up Business Unit of AREVA with Air liquide decided to start the development of a new technology based on the use of liquid nitrogen (−140°C / 3500 bar).

Nuclear operations with reactors and particle accelerators lead to the generation of neutrons, with a consequent activation of surrounding shielding materials. In general these materials consist of a concrete bioshield with associated metal work from steel reinforcement and support structures. This paper considers the interaction mechanisms of the neutrons in the concrete and the distribution and activity profile of activated materials. The most common source of high thermal neutron fluxes is of course from fission reactor systems. However, there are also other systems such as fusion reactors, which generate a far higher energy neutron flux, as the requirement for electrical confinement of the plasma means that the available bioshielding close to the power source is more limited. In addition, there are also neutron generation mechanisms from particle accelerator systems such synchrotrons and Van De Graaf devices. This study examines the interaction mechanisms of the neutrons in the bioshields, the radioelement abundances of the activated products generated, in-situ and ex-situ measurement of their concentration and profile of the radionuclides and finally methods of assaying and sentencing such materials. Examples of decommissioning operations of reactor and accelerator systems will be given as practical examples of the use of the assaying methodologies developed.