Around 250,000 tons of irradiated graphite (i-graphite) exists worldwide and can be considered as a current waste or future waste stream. The largest national i-graphite inventory is located in UK (∼ 100,000 tons) with significant quantities also in Russia and France [5]. Most of the i-graphite remains in the cores of shutdown nuclear reactors including the MAGNOX type in UK and the UNGG in France. Whilst there are still operational power reactors with graphite cores, such as the Russian RBMKs and the AGRs in UK, all of them will reach their end of life during the next two decades. The most common reference waste management option of i-graphite is a wet or dry retrieval of the graphite blocks from the reactor core and the grouting of these blocks in a container without further conditioning. This produces large waste package volumes because the encapsulation capacity of the grout is limited and large cavities in the graphite blocks could reduce the packing densities. Packing densities from 0.5 to 1 tons per cubic meter have been assumed for grouting solutions. Furthermore the grout is permeable. This could over time allow the penetration of aqueous phases into the waste block and a potential dissolution and release of radionuclides. As a result particularly highly soluble radionuclides may not be retained by the grout. Vitrification could present an alternative, however a similar waste package volume increase may be expected since the encapsulation capacity of glass is potentially similar to or worse than that of grout. FNAG has developed a process for the production of a graphite-glass composite material called Impermeable Graphite Matrix (IGM) [3]. This process is also applicable to irradiated graphite which allows the manufacturing of an impermeable material without volume increase. Crushed i-graphite is mixed with 20 vol.% of glass and then pressed under vacuum at an elevated temperature in an axial hot vacuum press (HVP). The obtained product has zero or negligible porosity and a water impermeable structure. Structural analysis shows that the glass in the composite has replaced the pores in the graphite structure. The typical pore volume of a graphite material is in the range of 20 vol.%. Therefore no volume increase will occur in comparison with the former graphite material. This IGM material will allow the encapsulation of i-graphite with package densities larger than 1.5 ton per cubic meter. Therefore a huge volume saving can be achieved by such an alternative encapsulation method. Disposal performance is also enhanced since little or no leaching of radionuclides is observed due to the impermeability of the material. NNL and FNAG have proved that IGM can be produced by hot isostatic pressing (HIP) which has several advantages for radioactive materials over the HVP process. • The sealed HIP container avoids the release of any radionuclides. • The outside of the waste package is not contaminated. • The HIP process time is shorter than the HVP process time. • The isostatic press avoids anisotropic density distributions. • Simple filling of the HIP container has advantages over the filling of an axial die.

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).

Processing liquid wastes frequently generates off gas streams with high humidity and liquid aerosols. Droplet laden air streams can be produced from tank mixing or sparging and processes such as reforming or evaporative volume reduction. Unfortunately these wet air streams represent a genuine threat to HEPA filters. High efficiency mist eliminators (HEME) are one option for removal of liquid aerosols with high dissolved or suspended solids content. HEMEs have been used extensively in industrial applications, however they have not seen widespread use in the nuclear industry. Filtering efficiency data along with loading curves are not readily available for these units and data that exist are not easily translated to operational parameters in liquid waste treatment plants. A specialized test stand has been developed to evaluate the performance of HEME elements under use conditions of a US DOE facility. HEME elements were tested at three volumetric flow rates using aerosols produced from an iron-rich waste surrogate. The challenge aerosol included submicron particles produced from Laskin nozzles and super micron particles produced from a hollow cone spray nozzle. Test conditions included ambient temperature and relative humidities greater than 95%. Data collected during testing HEME elements from three different manufacturers included volumetric flow rate, differential temperature across the filter housing, downstream relative humidity, and differential pressure (dP) across the filter element. Filter challenge was discontinued at three intermediate dPs and the filter to allow determining filter efficiency using dioctyl phthalate and then with dry surrogate aerosols. Filtering efficiencies of the clean HEME, the clean HEME loaded with water, and the HEME at maximum dP were also collected using the two test aerosols. Results of the testing included differential pressure vs. time loading curves for the nine elements tested along with the mass of moisture and solid material on each element at final dP. Plots of overall filtering efficiencies for DOP (spherical aerosol) and dry surrogate (aspherical aerosols) at specified dPs were computed for each filter. Filtering efficiencies were determined as a function of particle size. Curves were also generated showing the most penetrating particle size as a function of dP. A preliminary set of tests was conducted to evaluate spray location, duration, pressure, and wash volume for in-place cleaning the interior surface (reducing dP) of the HEME element. A variety of nozzle designs were evaluated and test results demonstrated the potential to overload the HEME (saturate filter medium) resulting in very high dPs and extensive drain times. At least one combination of spray nozzle design, spray location on the surface of the element, and spray time/pressure was successful in achieving extension of operational life.

The new batch pyrolysis plant in Studsvik is built primarily for treatment of uranium containing dry active waste, “DAW”. Several other waste types have been identified that are considered or assumed suitable for treatment in the pyrolysis plant because of the possibility to carefully control the atmosphere and temperature of the thermal treatment. These waste types must be characterised and an evaluation must be made with a BAT (Best Available Technology) perspective. Studsvik have performed or plan to perform lab scale pyrolysis tests on a number of different waste types. These include - Pyrophoric materials (uranium shavings) - Uranium chemicals that must be oxidised prior to being deposited in repository - Sludges and oil soaks (this category includes NORM-materials) - Ion exchange resins (both “free” and solidified/stabilised) - Bitumen solidified waste Methodology and assessment criteria for various waste types, together with results obtained for the lab scale tests that have been performed, are described.

Based on the decision of the State Office for Nuclear Safety, the Dukovany NPP has been obliged to secure the efficient capacities for the disposal of spent ion exchange resins. Therefore, in September 2010, based on the contract with supplier company AMEC Nuclear Slovakia s.r.o. has begun with pumping and treatment of ion exchange resins from the storage tank 0TW30B02, situated in the auxiliary building. The SIAL ® technology, developed in AMEC Nuclear Slovakia, has been used for the solidification purposes. This technology allows an on-site treatment of various special radioactive waste streams (resins, sludge, sludge/resins and borates) at the room temperature. The SIAL ® matrix and technology were licensed by the Czech State Office for Nuclear Safety in 2007. On-site treatment and solidification of spent ion exchange resins at Dukovany NPP involves process of resin removal from tank using remotely operated manipulator, resin transportation, resin separation from free water, resin filling into 200 dm 3 drums and solidification into SIAL ® matrix in 200 dm 3 drums using the FIZA S 200 facility. The final product is observed for compressive strength, leachability, radionuclide composition, dose rate, solids and total weight. After meeting the requirements for final disposal and consolidation, the drums are being transported for the final disposal to the Repository at Dukovany site. During the 3 month’s trial operation in 2010, and the normal operation in 2011 and 2012, 189 tons of dewatered resins have been treated into 1960 drums, with total activity higher than 920 GBq. At the end of trial run (2010), 22 tons of dewatered resins were treated into 235 drums. During standard operation approximately 91 tons in 960 drums (2011) and 76 tons in 765 drums (2012) were treated. The weights of resins in the drum ware in the range from 89–106 kg and compressive strength limit (10 MPa) has already been achieved 24 hours after fixation. The final measured strength values ranged from 19.0–34.7 MPa and real leachability values ranged from 0.03–0.65%, far below the 4% limit value. Collective effective dose of all workers in 2012 was 7.7 mSv (12.6 mSv in 2011, 6.2 mSv in 2010). Average individual effective dose in 2012 was 0.55 mSv (14 workers), and maximal individual effective dose was 2.25 mSv. This approach allows fast, safe and cost effective immobilization and transformation of dangerous radioactive waste such as sludge and resins into the solid form, which is suitable for long term storage or disposal.

An advanced method, in which electrodialysis separation of sodium nitrate and decomposition of nitrate ion are combined, has been developed to remove nitrate ion from low-level radioactive liquid wastes including nitrate salts of high concentration. An engineering scale apparatus with two electrodialytic devices, in which the sodium and nitrate ions were separately removed by each device, was produced on the basis of the results of fundamental investigation previously reported, and the performance of the apparatus was tested. Both the ions were successfully removed at the same time, though these ions were separately transferred using two electrodialytic devices. And also effect of several experimental parameters such as current and temperature on current efficiency of both the ions of each device was investigated.

The great amount of water used for cooling the stricken power reactors at Fukushima Dai-ichi following the earthquake and tsunami of 11 March 2011 has resulted in accumulation of cooling water so-called the remaining water in some buildings. As the cooling water is subsequently contaminated by fission products (FPs) and some other radioactive substances, it is necessary to decontaminate this ‘cooling water’ to reduce the volume of liquid radioactive waste and to reuse it again for cooling the affected reactors. Some methods are applied to remove the radioactive substances from the cooling water. However, after treatments of water, there arises a secondary radioactive waste, the sludge. Steps are now taken to immobilize this sludge. In this paper, BaSO 4 , as one of main constituents of the sludge, was chosen as an immobilizing target substance. The appropriate manufacturing condition of glass waste form for loading the sludge (BaSO 4 ) was studied and the chemical durability was evaluated by measuring the dissolution rate. For this experiment, the iron phosphate glass (IPG), which is known to possess a large loading capacity for a variety of chemical substances, was employed as the glass medium. Based on experimental results, it is evident that BaSO 4 can be loaded into the IPG medium when it possesses the appropriate composition and melting temperature. During loading, BaSO 4 converted into BaO, acting as a network modifier, which leads to enhanced stability of IPG.

Magnetite dissolution tests using a hydrazine base solution were performed at a temperature range of 90 to 150 °C. The dissolution rate of magnetite increased with [N 2 H 4 ], time, and temperature. The optimum solution pH in the experimental range was 3. The addition of copper ion to the hydrazine base solution greatly increased the magnetite dissolution rate. This was explained by the complex formation between N 2 H 4 and Cu ions, and the reducing power of the hydrazine-Cu complex to the ferric ions of magnetite. The reductive decontamination solution can be applied below 100 °C by the addition of copper ions. The chemical decontamination of a Type 304 stainless steel specimen using a hydrazine base reductive decontamination solution was also performed. The contact dose rate was greatly decreased by the repetitive application of NP and the hydrazine base solution.

The ability to monitor critical environment parameters of nuclear plants at all times, particularly during and after a disruptive accident, is vital for the safety of plant personnel, rescue and recovery crews, and the surrounding communities. Conventional hard-wired assets that depend on supplied power may be decimated as a result of such events, as witnessed in the Japanese Fukushima nuclear power plant in March 2011. Self-powered monitoring devices operating on a wireless platform, on the other hand, may survive such calamity and remain functional. The devices would be prepositioned at strategic locations, particularly where the dangerous build-up of contamination and radiation may preclude subsequent manned entrance and surveillance. Equipped with sensors for β-γ radiation, neutrons, hydrogen gas, temperature, humidity, pressure, and water level, as well as with criticality alarms and imaging equipment for heat, video, and other capabilities, these devices can provide vital surveillance information for assessing the extent of plant damage, mandating responses (e.g., evacuation before impending hydrogen explosion), and enabling overall safe and efficient recovery in a disaster. A radio frequency identification (RFID)-based system — called ARG-US — may be modified and adapted for this task. Developed by Argonne for DOE, ARG-US (meaning “watchful guardian”) has been used successfully to monitor and track sensitive nuclear materials packages at DOE sites. It utilizes sensors in the tags to continuously monitor the state of health of the packaging and promptly disseminates alarms to authorized users when any of the preset sensor thresholds is violated. By adding plant-specific monitoring sensors to the already strong sensor suite and adopting modular hardware, firmware, and software subsystems that are tailored for specific subsystems of a plant, a Remote Area Modular Monitoring (RAMM) system, built on a wireless sensor network (WSN) platform, is being developed by Argonne National Laboratory. ARG-US RAMM, powered by on-board battery, can sustain extended autonomous surveillance operation during and following an incident. The benefits could be invaluable to such critical facilities as nuclear power plants, research and test reactors, fuel cycle manufacturing centers, spent-fuel dry-cask storage facilities, and other nuclear installations.

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.

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.

Liquid low level radioactive effluents, when solidified in e. g. cement matrix, contribute to a significant extent to the waste amount to be disposed of in final repositories. Accordingly, since many years scientists and engineers investigate processes to remove the radioactive nuclides selectively from the effluents, to split the raw solution into two separate fractions: a large fraction with activity concentrations below the limits for free release; and a small fraction, containing the activity in concentrated form on e. g. ion exchanger materials (ion exchange has proven to be the most promising method for such “nuclide separation”). The challenge to be taken up is: When (and this is most often the case) the effluent contains organic materials and complexing agents, the formation of e. g. the 60-Co-EDTA complex prohibits its fixation to the ion exchangers. Accordingly the complexing agent needs to be removed or destroyed. The Institute for Chemistry of the Russian Academy of Sciences has applied the method of hydrothermal treatment (at elevated temperature and pressure, 200 °C, 200 bar), supported by Hydrogen Peroxide oxidation, to allow virtually complete removal of radioactive nuclides on inorganic ion exchangers. Pilot plants have been operated successfully in Russian power stations, and an operational plant has been designed. The method is being extended for an interesting and promising application: spent organic ion exchange resins, loaded up to the medium activity level, represent a serious disposal problem. With the hydrothermal process, in a process cycle, the activity can be stripped from the resins, the organic content is destroyed, and the activity is fixed on an inorganic absorber, well suited for final disposal.

Hydrogen generation mitigation for K Basin sludge was examined by encapsulation of uranium metal in BoroBond®, pre-oxidation of uranium metal with Fenton’s reagent and grinding of Densalloy SD170, an irradiated uranium metal surrogate. Encapsulation in BoroBond ® resulted in pressure increase rates at 60 °C ranging from 0.116 torr/h to 0.186 torr/h compared to 0.240 torr/h for a uranium metal in water standard. Samples cast with higher water content led to increased rates. A Fenton’s reagent system consisting of a simple reagent mix of FeSO 4 ·7H 2 O, H 2 O 2 and HCl effectively oxidized 1/4 ″ cubes of uranium metal in under four days at room temperature. Increased peroxide addition rate, increased FeSO 4 ·7H 2 O concentration and low pH all increase the corrosion rate. Densalloy SD170 with an average particle size of 581 μm with 7.63% of particles less than 90 μm was milled so that over 90% of the Densalloy mass measured less than 90 μm in 6 hours of milling. Acceptable wear rates were seen on wear components that were from standard materials (Nitronic SS and 440SS).

We developed a room temperature method to recondition intermediate-level homogeneous bituminized waste, called Eurobitum. Using this method, Eurobitum could be separated in an aqueous and an organic effluent, and a solid asphaltene-salt mix. In this paper, we present the results of the room temperature reconditioning method using real radioactive Eurobitum samples. The dissolution and separation processes were optimized. The residual waste could then be pressed towards a material, that might be suitable for geological disposal. The concentration of radionuclides in the two secondary waste effluents was thereafter determined. The aqueous effluent contained radionuclide concentrations below the treatment levels for common radioactive waste water treatment processes, whereas the alpha radioactivity of the organic phase was low enough to consider the possibility to inject it into the CILVA radwaste incinerator where it would undergo a thermal treatment process together with other solid radwaste. Our proposed process could reduce the Eurobitum volume considerably. The most flammable and the least radiation resistant molecules and the nitrates were removed from the original Eurobitum waste. Finally, the residual waste is expected to be more stable in geological disposal conditions.

Since 1979, Uranium enrichment technology has been researched through the gas centrifuge method, at Ningyo-toge Environmental Engineering Center of Japan Atomic Energy Agency (JAEA). In addition, the Demonstration Plant, that is final stage test facilities, was operating continuously from 1988 to 2001. As a result, a lot of residues accumulated in the plant. Most of this accumulation was found be uranium intermediate fluoride. The basic decommission policy of JAEA is that equipments of gas centrifuge will be decontaminated by sulfuric acid immersion method for clearance and reuse. In our plan, approximately 90% of metals will be cleared and reused, and then the remaining 10% will be disposed of radioactive waste. We propose a combination of sulfuric acid immersion method and the systematic chemical decontamination as an efficient method for decontamination of uranium enrichment facilities. This paper focuses on the method and performance of systematic chemical decontamination using IF 7 gas. The following (Figure 1) shows our decommission policy and position of systematic chemical decontamination by IF 7 gas for uranium enrichment plant. The IF 7 treatment technique belongs to the systematic decontamination technology. It has the high performance decontamination technique for the plant that accumulates the uranium intermediate fluoride, such as UF 4 , UF 5 , U 2 F 9 , and U 4 F 17 , which exist in the uranium enrichment plant through the Gas Centrifuge, called GCF. The one of characteristics of the IF 7 treatment, the secondary waste is just an IF 5 and little residues. In addition, this IF 5 can be reused as materials for making new IF 7 gas. The IF 7 treatment can also be performed in the room temperature and very low pressure like a 10–45hPa. Furthermore, the IF 7 treatment is a simple method using chemical reaction. For this reason, we hardly need to care about secondary reaction with the exception of the reaction with IF 7 gas and the uranium intermediate fluoride. This is a very important feature when applying to a large-scale plant. In order to carry out the IF 7 treatment, we only set up a few equipments in GCF uranium enrichment plant, which were IF 7 feeding equipment and two circulating pumps. IF 7 gas cylinders are seated in IF 7 feeding equipment. This is the only equipment. Figure 2 shows the IF 7 treatment system. We carried out the IF 7 treatment for the four cascades in the uranium enrichment Demonstration Plant. The weights of uranium residue in the cascades were approximately 700kgU per cascade prior to the IF 7 treatment. In the IF 7 treatment, we were able to find the near-optimal processing condition. As a result, we could confirm the IF 7 treatment period for one cascade which was 60 days. The main factor to determine the IF 7 treatment period is the pressure and the flow rate of reaction product gas (UF 6 and IF 5 mixture gas) exhausted from the cascade. Although we carried out the IF 7 treatment with the maximum value of the flow rate, which our facility has, it is possible to further shorten the IF 7 treatment period by setting a higher gas flow rate. Moreover, after the IF 7 treatment, we evaluated the uranium recovery rate for cascades and the residues’ uranium weight in the main equipment of GCF. In addition, in the evaluation of the uranium recovery rate, we enable to confirm the uranium recovery rate of all cascades achieved more tan 98%. Furthermore, the average of uranium recovery rate more than 99% in the cascade that has been processed at the end. As a result, radioactive concentration of uranium in the main parts of the GCF fell to 1.0B q/g and below.

Life-limiting features of stores (i.e. components of a typical Intermediate Level Waste (ILW) store that are vulnerable to failure) can be separated into those related to store design, the continuing operation of the store and equipment and store management. Environmental factors, including temperature, humidity, atmospheric pollutants and radiation fields also affect the longevity of ILW storage buildings and the integrity of the waste packages within the store. The current condition of the store and the causes and rates of any observed deterioration need to be determined to estimate the current operating life of the store and to analyse the costs and benefits associated with life extension. Suitable monitoring techniques and methods for extending the operating life of stores, either by remediation or replacement of identified life-limiting features, have also been assessed. Techniques including visual inspection (directly or by use of CCTV or inspection cells), corrosion coupons, crane monitoring and a wide range of indirect measurement methods can be used to monitor the condition of life-limiting features and the store environment. Methods for extending store life include the application of paint or coatings to retard degradation, use of electro-chemical treatments and repair of the concrete structure or replacement of components.

This paper deals with information on the radioactive waste cementation technology for decommissioning of Salaspils Research Reactor (SRR). Dismantled and segmented radioactive materials were cemented in concrete containers using tritiated water-cement mixture. The viscosity of water-cement mortar, mechanical tests of solidified mortar’s samples, change of temperature of the samples during solidification time and long time leakage of 137 Cs, 14 C, 60 Co and 3 T radionuclides was studied for different water-cement compositions with additives. The pH and electro conductivity of the solutions during leakage tests were controlled. It was shown, that water/cement ratio significantly influences on water-cement mortar’s viscosity and solidified samples mechanical stability. The role of additives — fly ash and Penetron admix in reduction of solidification temperature is discussed. It was found, that addition of fly ash to the cement-water mortar can reduce the solidification temperature from 81°C up to 62°C. The optimal interval of water ratio in cement mortar is discussed. Radionuclides leakage tests show that the release curves has a complicate structure. The possible radionuclides release mechanisms are discussed. Experimental results indicated that additives can significantly influence on the radionuclides release processes from cemented samples. The optimization of cementation of radioactive wastes in concrete containers was performed using mechanical stability, solidification temperature, radionuclide releases and viscosity of mortar.

The IAEA Coordinated Research Project (CRP) on cementitious materials for radioactive waste management was launched in 2007 [1, 2]. The objective of CRP was to investigate the behaviour and performance of cementitious materials used in radioactive waste management system with various purposes and included waste packages, wasteforms and backfills as well as investigation of interactions and interdependencies of these individual elements during long term storage and disposal. The specific research topics considered were: (i) cementitious materials for radioactive waste packaging: including radioactive waste immobilization into a solid waste form, (ii) waste backfilling and containers; (iii) emerging and alternative cementitious systems; (iv) physical-chemical processes occurring during the hydration and ageing of cement matrices and their influence on the cement matrix quality; (v) methods of production of cementitious materials for: immobilization into wasteform, backfills and containers; (vi) conditions envisaged in the disposal environment for packages (physical and chemical conditions, temperature variations, groundwater, radiation fields); (vii) testing and nondestructive monitoring techniques for quality assurance of cementitious materials; (viii) waste acceptance criteria for waste packages, waste forms and backfills; transport, long term storage and disposal requirements; and finally (ix) modelling or simulation of long term behaviours of cementations materials used for packaging, waste immobilization and backfilling, especially in the post-closure phase. The CRP has gathered overall 26 research organizations from 22 Member States aiming to share their research and practices on the use of cementitious materials [2]. The main research outcomes of the CRP were summarized in a summary report currently under preparation to be published by IAEA. The generic topical sections covered by report are: a) conventional cementitious systems; b) novel cementitious materials and technologies; c) testing and waste acceptance criteria; and d) modelling long term behaviour. These themes as well as all contributions of participating organizations were further developed in the individual reports to be presented in the IAEA publication. The CRP facilitated the exchange of information and research co-operation in resolving similar problems between different institutions and contributed towards improving waste management practices, their efficiency and general enhancement of safety.

The evolution of the engineered barrier system (EBS) of geological repositories for radioactive waste has been the subject of many research programmes during the last decade. The emphasis of the research activities was on the elaboration of a detailed understanding of the complex thermo-hydro-mechanical-chemical processes, which are expected to evolve in the early post closure period in the near field. It is important to understand the coupled THM-C processes and their evolution occurring in the EBS during the early post-closure phase so it can be confirmed that the safety functions will be fulfilled. Especially, it needs to be ensured that interactions during the resaturation phase (heat pulse, gas generation, non-uniform water uptake from the host rock) do not affect the performance of the EBS in terms of its safety-relevant parameters (e.g. swelling pressure, hydraulic conductivity, diffusivity). The 7th Framework PEBS project (Long Term Performance of Engineered Barrier Systems) aims at providing in depth process understanding for constraining the conceptual and parametric uncertainties in the context of long-term safety assessment. As part of the PEBS project a series of laboratory and URL experiments are envisaged to describe the EBS behaviour after repository closure when resaturation is taking place. In this paper the very early post-closure period is targeted when the EBS is subjected to high temperatures and unsaturated conditions with a low but increasing moisture content. So far the detailed thermo-hydraulic behaviour of a bentonite EBS in a clay host rock has not been evaluated at a large scale in response to temperatures of up to 140°C at the canister surface, produced by HLW (and spent fuel), as anticipated in some of the designs considered. Furthermore, earlier THM experiments have shown that upscaling of thermal conductivity and its dependency on water content and/or humidity from the laboratory scale to a field scale needs further attention. This early post-closure thermal behaviour will be elucidated by the HE-E experiment, a 1:2 scale heating experiment setup at the Mont Terri rock laboratory, that started in June 2011. It will characterise in detail the thermal conductivity at a large scale in both pure bentonite as well as a bentonite-sand mixture, and in the Opalinus Clay host rock. The HE-E experiment is especially designed as a model validation experiment at the large scale and a modelling programme was launched in parallel to the different experimental steps. Scoping calculations were run to help the experimental design and prediction exercises taking the final design into account are foreseen. Calibration and prediction/validation will follow making use of the obtained THM dataset. This benchmarking of THM process models and codes should enhance confidence in the predictive capability of the recently developed numerical tools. It is the ultimate aim to be able to extrapolate the key parameters that might influence the fulfilment of the safety functions defined for the long term steady state.

It removes radioactive sludge and corrosion products deposited on the inner walls of the pipes and valves in replacement or decommission, upon termination of life time, of nuclear power plant or nuclear facility. It lowers a cost of waste treatments taking advantage of a reduction of quantity of radioactive wastes by treating in classification of the radioactive wastes whose activities are lower than legal standards. The cleaning or decontamination methods developed until now have induced a damage on systems while being operated. A decontamination has been restrained if it was difficult to access physically. We are in development of the cleaning technique for pipelines by utilizing micro-bubbles in order to improve an efficiency and to prevent from any damage of systems. It aims to conduct a decontamination for spaces difficult to access there by applying cavitation phenomenon that is generated in collapse of micro-bubbles. In order to improve an efficiency of the micro-bubble device, the experimental conditions suitable to decontamination have been established and the auxiliary equipments have been added. The generation conditions and characteristics of micro-bubbles have been demonstrated by adjusting pHs and temperatures of cleaning solution. A decontamination effect has been confirmed by adding up an electrolytic method and ozone into micro-bubbles.