In 2008, following the general French plan for nuclear waste management, Électricité de France attempted to find for irradiated graphite an alternative solution to direct storage at the low-activity long-life storage center in France managed by the national agency for wastes (ANDRA). EDF management requested that its engineering arm, EDF CIDEN, study the graphite treatment alternatives to direct storage. In mid-2008, this study revealed the potential advantage for EDF to use a steam reforming process known as THermal Organic Reduction, “THOR” (owned by Studsvik, Inc., USA), to treat or destroy the graphite matrix and limit the quantity of secondary waste to be stored. In late 2009, EDF began a test program with Studsvik to determine if the THOR steam reforming process could be used to destroy the graphite. The program also sought to determine if the graphite could be treated to release the bulk of activity while minimizing the gasification of the bulk mass of the graphite. In October 2009, tests with non-irradiated graphite were completed and demonstrated destruction of a graphite matrix by the THOR process at satisfactory rates. After gasifying the graphite, focus shifted to the effect of roasting graphite at high temperatures in inert gases with low concentrations of oxidizing gases to preferentially remove volatile radionuclides while minimizing the graphite mass loss to 5%. A radioactive graphite sleeve was imported from France to the US for these tests. Completed in April 2010, ‘Phase I’ of testing showed that the process removed >99% of H-3 and 46% of C-14 with <6% mass loss. Completed in September 2011, ‘Phase II’ testing achieved increased removals as high as 80% C-14. During Phase II, it was also discovered that roasting in a reducing atmosphere helped to limit the oxidation of the graphite. Future work seeks to explore the effects of reducing gases to limit the bulk oxidation of graphite. If the graphite could be decontaminated of long-lived radionuclides up to 95% for C-14 while minimizing mass loss to <5%, this would minimize the volume of any secondary waste streams and potentially lower the waste class of the larger bulk of graphite. Alternatively, if up to 95% decontamination of C-14 is achieved, the graphite may be completely gasified which could result in lower disposal volumes.

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.

In nuclear power plants (NPP) ion exchange (IX) resins are used in several systems for water treatment. The resins can be in bead or powdered form. For waste treatment of spent IX resins, two methods are basically used: • Direct immobilization (e.g. with cement, bitumen, polymer or High Integrity Container (HIC)) • Thermal treatment (e.g. drying, oxidation or pyrolysis) Bead resins have some properties (e.g. particle size and density) that can have negative impacts on following waste treatment processes. Negative impacts could be: • Floatation of bead resins in cementation process • Sedimentation in pipeline during transportation • Poor compaction properties for Hot Resin Supercompaction (HRSC) Reducing the particle size of the bead resins can have beneficial effects enhancing further treatment processes and overcoming prior mentioned effects. Westinghouse Electric Company has developed a modular grinding process to crush/grind the bead resins. This modular process is designed for flexible use and enables a selective adjustment of particle size to tailor the grinding system to the customer needs. The system can be equipped with a crusher integrated in the process tank and if necessary a colloid mill. The crusher reduces the bead resins particle size and converts the bead resins to a pump able suspension with lower sedimentation properties. With the colloid mill the resins can be ground to a powder. Compared to existing grinding systems this equipment is designed to minimize radiation exposure of the worker during operation and maintenance. Using the crushed and/or ground bead resins has several beneficial effects like facilitating cementation process and recipe development, enhancing oxidation of resins, improving the Hot Resin Supercompaction volume reduction performance.

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.

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).

The decontamination factor of liquid waste containing 60 Co is generally weak. This is due to the presence of complexant molecules. For instance, complexation of EDTA with 60 Co decreases efficiency of radioactive waste treatment. The aim of this study was to degrade EDTA in H 2 O and CO 2 and to concentrate free 60 Co in order to increase decontamination factor. A first test of radioactive waste treatment by photocatalysis was allowed to increase decontamination factor ( 60 Co) from 16 to 196 with a device requiring to be improved. The present work concerns the first step of the degradation process development with a more powerful device. These first experiments were leaded to follow the only EDTA oxidation. EDTA degradation was carried out by the following Advanced Oxidation Processes (AOP): UV/H 2 O 2 (photochemistry); UV/TiO 2 (photocatalysis); UV/TiO 2 /H 2 O 2 . A specific reactor was achieved for this study. The wavelength used was 254 nm (UVC). The photocatalytic degradation of EDTA was carried out with Degussa P-25 titanium dioxide (TiO 2 ), which is a semiconductor photocatalyst. The degradation degree of EDTA and the intermediate products were monitored by TOC and ionic chromatography methods. The effects of various parameters such as pH and the quantity of H 2 O 2 were studied. This allows us to conclude that basic pH slows down EDTA degradation. The study showed that UV/H 2 O 2 process was the most effective treatment process under acid conditions. The rate of EDTA degradation was very high and reached 95% in 120 minutes. The presence of glyoxilic, oxalic, glycolic and formic acids was detected as degradation products. Among the intermediates produced by photochemistry, NO 3 − ions presence informed of the amine degradations. These results highlighted faster EDTA degradation by photochemistry than photocatalysis.

Decontamination prior to decommissioning and dismantling is an internationally accepted approach. Not only does it provide for minimization of personnel dose exposure but also maximization of the material volume available for free release. Since easier dismantling techniques in lower dose areas can be applied, the licensing process is facilitated and the scheduling and budgeting effort is more reliable. The most internationally accepted approach for decontamination prior to decommissioning projects is the Full System Decontamination (FSD). FSD is defined as the chemical decontamination of the primary cooling circuit, in conjunction with the main auxiliary systems. AREVA has long-term experience with Full System Decontamination for return to service of operating nuclear power plants as well as for decommissioning after shutdown. Since 1976, AREVA has performed over 500 decontamination applications and from 1986 on, decontaminations prior to decommissioning projects which comprise virtually all nuclear power plant (NPP) designs and plant conditions: • NPP designs: HPWR, PWR, and BWR by AREVA, Westinghouse, ABB and GE; • Decontaminations performed shortly after final shutdown or several years later, and even after re-opening safe enclosure; • High alpha inventory and or low gamma/alpha ratio; • Main coolant chemistry (e.g. with and without Zn injection during operation). Fifteen decontaminations prior to decommissioning projects have been performed successfully to date. The lessons learned of each project were consequently implemented for the next project. AREVA NP has developed a fully comprehensive approach for decontamination based on the CORD® (Chemical Oxidation Reduction Decontamination) Family, applied using the in-house designed decontamination equipment AMDA® (Automatic Modular Decontamination Appliance). The Decontamination Concept for Decommissioning (DCD) will be outlined in this paper. Based on highlights of previous FSDs performed prior to decommissioning the AREVA concept for FSD and DCD will be outlined: • Application window; • Decontamination area; • Waste considerations; • Positive results for subsequent decommissioning and dismantling activities.

Disposal of organic liquid waste has become an increasing issue for many nuclear sites. Existing disposal solutions such as incineration or super critical water techniques are not compatible with wastes containing chlorinated solvents or fluorine owing to corrosion problems. As an example several hundred cubic meters of lubricating oils mixed with trichloroethylene (TCE) or perchloroethylene (PCE) are stockpiled on several French nuclear sites. For several years Dewdrops has been developing an original combination of mineralization processes for waste oils and solvents particularly well suited to the nuclear field. The patented technology relies on the alternation of chemical and biological oxidation mechanisms. The oxidized organic material predominately forms carbon dioxide, water and inorganic salts. This paper details the procedure and the results obtained for a particular case at the Tricastin nuclear site of Areva NC (South France). The organic waste used in this study was a 85/15 v/v ratio mix of lubricating oil and TCE. The pilot plant build upon the technology has a daily treatment capacity of approximately 10 liters. In the first step the TCE is mineralized by the photo Fenton reaction. Using hydrogen peroxide with an ultraviolet regenerated iron catalyst, TCE is transformed to carbon dioxide, water and hydrochloric acid. After neutralizing with caustic soda, the next step is a multi-stage biodegradation process to eliminate the remaining lubricating oil. Carefully selected microorganisms use the organic waste as an energy source for their metabolism. During oil biodegradation over 75% of the carbon is released as carbon dioxide while the remaining is incorporated into the biomass. The aqueous phase is continuously separated from the biomass using cross flow filters. The output aqueous phase is treated with ozone and ultraviolet light to eliminate the remaining organic compounds. The final effluent obtained is in conformance with European water standards and can be disposed by normal means. It can also be adjusted to local requirements. The radioactive elements and heavy metals present in these lubricating oils are trapped by the biomass. The excess of biomass is recovered by centrifugation and mineralized by catalytic ozonation technology. The result of the tests was a radioactive waste reduction factor of 15. The mineral residue is a concentrate of inorganic salts with traces of radioactive elements as well as heavy metals. The radioactive elements thus recovered can be consigned to an official repository.

Part of TRU waste includes a large amount of nitrate salt, the effects of which have to be evaluated in a safety assessment of co-location disposal with high level radioactive waste (HLW). High concentrations of nitrate ions from TRU waste might affect the solubility of different radionuclides in the HLW. In the current study, the effects of nitrate salt on radionuclide solubility were investigated experimentally. Solubility experiments of important and redox sensitive radionuclides, Tc(IV), Np(IV) and Se(0), were performed using various concentrations of sodium nitrate (NaNO 3 ) and of Np(V) in NaNO 3 solutions to investigate complex formation with NO 3 − ions. Solubility experiments of Pd(II), Sn(IV) and Nb(V) using ammonium chloride (NH 4 Cl) solution were also undertaken to investigate complex formation with NH 3 /NH 4 + ions. No significant solubility enhancement was observed for Np and Se. Tc solubility in ≥0.1 mol/dm 3 NaNO 3 solution increased due to oxidation by nitrate ions. An increase of Np(V) solubility was expected by the chemical equilibrium model calculation with JNC-TDB, however, solubility enhancement by complex formation of Np(V) with nitrate ions was not observed. Solubility enhancement by complex formation of Sn and Nb were also not observed, only Pd solubility was increased by complex formation with NH 3 /NH 4 + ions.

Chemical state of elements in groundwater is one of the most important information for understanding behavior of elements in underground environment. Chemical state of elements controlled mainly by groundwater physico-chemical parameters. Because the change of physico-chemical parameters of groundwater, due to pressure release and oxidation during sampling, causes changes in chemical state of elements, systematic methodologies for understanding in situ chemical state is required. In this study, in order to understand chemical state of elements in groundwater, an ultrafiltration instrument for maintaining in-situ pressure and anaerobic conditions was developed. The instrument developed in this study for ultrafiltration made of passivated Stainless Used Steel (SUS) materials, was designed to keep groundwater samples maintaining in-situ pressure/anaerobic conditions. Ultrafiltration of groundwater was conducted at a borehole drilled from the 200 mbGL (meters below ground level) Sub-stage at a depth of 200 m at the Mizunami Underground Research Laboratory. Chemical analyses of groundwater were also conducted using samples filtered under both pressurized/anaerobic and atmospheric conditions and passivated SUS materials with different elapsed times after passivation. The results indicate that our ultrafiltration method is suitable for collection of filtered groundwater and passivation is an essential treatment before ultrafiltration.

The radioactivity of 3 H of the metal samples from the nuclear sites was analyzed by using a commercialized high temperature furnace and a Liquid Scintillation Counter (LSC). The 3 H activity of the sample was measured according to the duration of the high temperature combustion and the oxidation temperature. Basically, the recovery from the furnace was 90% for 3 H and the LSC had a quenching efficiency of approximately 30 %. HNO 3 was used as a trapping solution for 3 H and the solution was cocktailed with a scintillator. The activity extracted from the sample was increased till the combustion time elapsed 60 minutes and the increasing rate was reduced continuously thereafter at 600 °C whereas 80% of radioactivity was extracted during the first 15 minutes at 900 °C. Also, the pretreatment for the metal sample, which included a high temperature combustion and trapping, had the time required of at least four hours at 900 °C. Finally, it was suggested that this high temperature combustion method could be applied to analyze the activity of the radioactive metal waste from the nuclear power plants.

In the Korean LILW (Low- and Intermediate-Level radioactive Waste) repository at Gyeongju city, the degradation of organic wastes and the corrosion of metallic wastes and steel containers would be important processes that affect repository geochemistry, speciation and transport of radionuclides during the lifetime of a radioactive waste disposal facility. Gas is generated in association with these processes and has the potential threat to pressurize the repository, which can promote the transport of groundwater and gas, and consequently radionuclide transport. Microbial activity plays an important role in organic degradation, corrosion and gas generation through the mediation of reduction-oxidation reactions. The Korean research project on gas generation is being performed by Korea Radioactive Waste Management Corporation (hereafter referred to as “KRMC”). A full-scale in-situ experiment will form a central part of the project, where gas generation in real radioactive low-level maintenance waste from nuclear power plants will be done as an in-depth study during ten years at least. In order to examine gas generation issues from an LILW repository which is being constructed and will be completed by the end of December, 2012, two large-scale facilities for the gas generation experiment will be established, each equipped with a concrete container carrying on 16 drums of 200 L and 9 drums of 320 L of LILW from Korean nuclear power plants. Each container will be enclosed within a gas-tight and acid-proof steel tank. The experiment facility will be fully filled with ground water that provides representative geochemical conditions and microbial inoculation in the near field of repository. In the experiment, the design includes long-term monitoring and analyses for the rate and composition of gas generated, and aqueous geochemistry and microbe populations present at various locations through on-line analyzers and manual periodical sampling. A main schedule for establishing the experiment facility is as follows: Completion of the detailed design until the second quarter of the year 2010; Completion of the manufacture and on-site installation until the second quarter of the year 2011; Start of the operation and monitoring from the third quarter of the year 2011.

Fuji Electric had developed the low pressure oxygen plasma technology for mild decomposition and mineralization of an organic material such as ion exchange resin. This method is suitable for radioactive spent resin volume/weight reduction and stabilization for final disposal. On this process, the ion-exchange resins are vaporized and decomposed into gas-phase with pyrolysis, and then, they are decomposed and oxidized with low-pressure plasma activity based on oxygen. And this process is achieved under moderate condition for radio active waste. • incinerate temperature: 400–700 deg C; • low-pressure (low-temperature) plasma condition: 10–50 Pa. From the result of this process, named of LPOP(low pressure oxidation process) by the inductively coupled plasma, we have confirmed that the process is applicable for organic fireproof waste including ion-exchange resin, and found that the used resin treatment performance is the same as cold test (using imitate spent resin) [1] [2] [3]. In this paper, the outline of the LPOP technology, and two research results on the possibility of solidification with cement of LPOP residue for geological disposes are reported. (1)Study of the residue chemical form after LPOP process (2)Study of the solidification character with cement.

Decontamination prior to Decommissioning and Dismantlement is imperative. Not only does it provide for minimization of personnel dose exposure but also maximization of the material volume available for free release. Since easier dismantling techniques in lower dose areas can be applied, the licensing process is facilitated and the scheduling and budgeting effort is more reliable. The most internationally accepted approach for Decontamination prior to Decommissioning projects is the Full System Decontamination (FSD). FSD is defined as the chemical decontamination of the primary cooling circuit, in conjunction with the main auxiliary systems. AREVA NP has long-term experience with Full System Decontamination for return to service of operating nuclear power plants as well as for decommissioning after shutdown. Since 1976, AREVA NP has performed over 500 decontamination applications and, from 1986, Decontaminations prior to Decommissioning projects which comprise virtually all NPP designs and plant conditions were performed: • NPP designs: HPWR, PWR, and BWR by AREVA, Westinghouse, ABB and GE; • Decontaminations performed shortly after final shutdown or several years later, and even after re-opening Safe Enclosure; • High Alpha inventory and or low gamma/alpha ratio; • Main Coolant chemistry (e.g., with and without Zn injection during operation). Fifteen Decontaminations prior to Decommissioning Projects have been performed successfully to date and the sixteenth FSD is now in the detailed engineering phase and is scheduled to commence late 2010. AREVA NP has developed a fully comprehensive approach for decontamination based on the CORD® (Chemical Oxidation Reduction Decontamination) Family, applied using the in-house designed decontamination equipment AMDA™ (Automatic Modular Decontamination Appliance). Based on the vast experience of AREVA NP in the field of decontamination, the Decontamination Concept for Decommissioning was developed. This concept ensures that the decontamination is tailored to the given boundaries and desired goals to ensure the best results with the lowest waste generation. This includes lower source term by oxide film, thus corrosion product, removal; controlled base metal attack to remove embedded activity; increased gamma/alpha ratios; and alpha contamination removal. This paper will describe the AREVA NP Decontamination Concept for Decommissioning (DCD) and present highlights of previous FSDs performed prior to decommissioning using the CORD/AMDA technology.

One of the main problems of liquid radioactive waste (LRW) management is concerned with treatment of decontamination waters containing organic ligands. The organic ligands like oxalic, citric and ethylenediaminetetraacetic acids form stable complexes with radionuclides which puts restrictions on application of many technologies of LRW management. One of the ways of destruction of metallorganic complexes consists in using the catalytic oxidation. However, the heterophase catalytic oxidation is rather problematic due to formation of metal oxides on the catalyst surface and calmatation of meso- and micropores. A possible solution of the above problem can be found in synthesis of macroporous catalysts for oxidation having a regular macroporous structure. The present paper describes the template synthesis of macroporous metalloxide catalysts performed with using siloxane-acrylate microemulsions as templates. The method for impregnation of precious metals (PM) particles into the template, which enables one to produce PM nanoparticles of a specific size and immobilize them in the porous structure of the synthesized metalloxide catalysts, is presented. A possible mechanism of the synthesis of macroporous catalysts is suggested and the comparison of the electronic and photon-correlation spectroscopy results obtained at different stages of catalysts synthesis was conducted.

Currently in the nuclear industry, surface contamination in the form of radioactive metal or metal oxide deposits is most commonly removed by chemical decontamination, electrochemical decontamination or physical attrition. Physical attrition techniques are generally used on structural materials (concrete, plaster), with (electro)chemical methods being used to decontaminate metallic or painted surfaces. The most common types of (electro)chemical decontamination are the use of simple mineral acids such as nitric acid or cerium (IV) oxidation (MEDOC). Use of both of these reagents frequently results in the dissolution of a layer of the substrate surface increasing the percentage of secondary waste which leads to burdens on downstream effluent treatment and waste management plants. In this context, both mineral acids and MEDOC can be indiscriminate in the surfaces attacked during deployment, e.g. attacking in transit through a pipe system to the site of contamination resulting in both diminished effect of the decontaminating reagent upon arrival at its target site and an increased secondary waste management requirement. This provides two main requirements for a more ideal decontamination reagent: Improved area specificity and a dissolution power equal to or greater than the previously mentioned current decontaminants. Photochemically promoted processes may provide such a decontamination technique. Photochemical reduction of metal ion valence states to aid in heavy metal deposition has already been extensively studied [1], with reductive manipulation also being achieved with uranium and plutonium simulants (Ce) [2]. Importantly photooxidation of a variety of metals, including neptunium [3], has also been achieved. Here we report on the potential application of this technology to metal dissolution.

Co-precipitation experiments of lanthanides were carried out using this lab-scale apparatus (4kg-salt/batch). As lanthanides, 8 lanthanide elements (Y, La, Ce, Pr, Nd, Sm, Eu and Gd) were used. By reaction with oxygen, these 8 lanthanide chlorides were converted their oxide (REO 2 , RE 2 O 3 ) or oxychloride form. Since these lanthanide oxides or oxychlorides are nearly molten salt insoluble, they all were precipitated by free settling in the bottom of molten salt bed, where about 7–8 hrs precipitation time was requested. It was found that in the conditions of 700 °C - 12 hours sparging time and 5 L/min, all the used lanthanide elements showed over 99.5% oxidation efficiency. But in case of 800 °C molten salt temperature only after 7 hours they showed over 99% oxidation efficiency.

Different models describing the dissolution mechanism of spent nuclear fuel under repository conditions have been developed in the last years. One of the most evolved ones is the Matrix Alteration Model (MAM), which is an Alteration/Dissolution source term model based on the oxidative dissolution of spent fuel. Oxidant and reducing species can be naturally or radiolytically-generated. The experimentally-observed inhibition of matrix dissolution by H 2 , was integrated into MAM by considering that H 2 is able to consume the oxidant species responsible for UO 2 oxidation, e.g. H 2 O 2 . As a consequence, MAM predicts lower H 2 O 2 concentrations for systems containing larger amounts of dissolved H 2 . Radiolysis experiments carried out by Pastina and coworkers have shown that under specific conditions such as high linear energy transfer (LET) radiation and absence of solid phase, dissolved H 2 has a negligible effect on the H 2 O 2 concentration, thus suggesting that the H 2 inhibition effect catalyzed by the matrix surface has not been properly implemented in MAM. Modelling exercises performed in this work confirm such point and reveal the necessity of considering H 2 -activation when modelling this kind of systems. In addition, it has been demonstrated that, for high LET radiation, a clear dependence exists between the extent of the H2 activation and the integral LET of the radiation. The integration of the function describing such dependence allows improving the implementation of the H 2 inhibition effect in MAM.

A novel technology was developed for treatment of evaporator concentrates produced as a result of operation of evaporation devices comprising the main component of special water purification systems of nuclear power plants (NPP). The developed technology includes a hydrothermal (T = 250–300°C and P = 80–120 bar) processing of evaporator concentrates in oxidation medium in order to destruct stable organic complexes of cobalt radionuclides and remove these radionuclides by oxide materials formed during such a processing. The cesium radionuclides contained in evaporator concentrates are removed by a conventional method — through application of one of the developed composite sorbents with ferrocyanides of transition metals used as active agents. Extensive laboratory studies of the processes occurring in evaporator concentrates under hydrothermal conditions were performed. It was shown that hydrothermal oxidation of evaporator concentrates has a number of advantages as compared to traditional oxidation methods (ozonation, photocatalytic, electrochemical and plasma oxidation). A laboratory installation was built for the flow-type hydrothermal oxidation of NPP evaporator concentrates. The obtained experimental results showed good prospects for the developed method application. On the basis of the results obtained, a pilot installation of productivity up to 15 l/hour was developed and built in order to work out the technology of evaporator concentrates hydrothermal treatment. The pilot tests of the hydrothermal technology for evaporator concentrates hydrothermal treatment were performed for 6 months in 2006 at the 1 st reactor unit of the Novovoronezhskaya NPP (Voronezh Region, Russia). Optimal technological regimes were determined, and estimations of the economic soundness of the technology were made. The advantages of the presented technology in terms of management of concentrated liquid radioactive wastes (LRW) at nuclear cycle facilities, as compared to other methods applicable for this type of LRW, were demonstrated. Application of the hydrothermal technology in the system of NPP LRW management enables one to reduce substantially the volume of solid radioactive waste sent for final disposal.