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.

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).

Diamond tools are well proven cutting, drilling and grinding technologies in many applications but need to be specifically optimized and adapted for the complex and varied structures of nuclear power plant in view of decontamination and decommissioning. The proper development and use of diamond tools in these extreme and complex conditions can only be achieved thanks to the combined talent of experienced nuclear plant contractors, engineers, technicians, operators of diamond tools, and the use of specialized equipment. This present paper is an overview of the possible applications of diamond tools in the different operations of Nuclear Decommissioning and Decontamination.

A decommissioning project for a uranium conversion plant was conducted to restore it to a safe environmental condition and minimal low level radioactive wastes which were converted to stable chemical forms for a long term disposal. In the middle of 2004, a decommissioning program for a conversion plant, which was constructed in 1982, and treated about 300 tons of natural uranium until it was shut down in 1992, obtained its approval from the regulatory body. Actual dismantling and decontaminating activities have been performed since July 2004 and will be finished by December 2009. The decommissioning works were mainly divided into two parts: the inside of the building containing the process equipments; the lagoon sludge generated during the plant operation. The decommissioning of the inside of the building was carried out by dismantling the process equipment, which were firstly segmented and decontaminated by polishing and washing with steam and chemicals or melting, and then decontamination for the surfaces inside the building by excavating or grinding the concrete walls. The decontamination goals were below 0.2Bq/g for the metallic segments and below 0.4Bq/cm 2 for the concrete walls. Decontamination methods were selected according to the degree of contamination and a minimization of the low level radioactive wastes was conducted throughout the decommissioning work. The lagoon sludge waste had two types, one was an various inorganic nitrate salt mixture containing a very low concentration of uranium, about 200∼300ppm, in Lagoon-II and the other was an inorganic nitrate salt mixture containing a few percent of uranium in Lagoon-I. To treat these sludge wastes a thermal decomposition facility was constructed and operated to produce stable sludge wastes containing uranium oxides which are stable in the air. The final sludge wastes after a thermal treating for the sludge waste of Lagoon-I could be reused. The final residual radioactivity for the inside of the building will be measured to confirm a complete decontamination of the uranium to back ground level and then the building will be considered for another use.

By contract with the Austrian government, the ARC is treating radioactive waste from research institutions and industries. In the last years, one focus was the development of processes for the treatment of NORM and TENORM. Our goal in developing such processes is to recycle valuable compounds for further industrial usage and to concentrate the radioactive elements as far as possible, to save space in the waste storage facilities. Austria is an important producer of tungsten-thoria- and tungsten-molybdenum-thoria-cermets. Scrap is generated during the production process in the form of turnings and grinding sludge and dust. Although big efforts have been undertaken to replace Thorium compounds, waste streams from past production processes are still waiting for treatment. The total amount of this waste stored in Austria may be estimated to be approx. 100 tons. In close co-operation with the tungsten industries, recycling processes were tested and further developed at ARC in laboratory, bench scale and pilot plants. Three different approaches to solve the problem were studied: Dissolution of tungsten in molten iron in an arc or induction furnace, thus producing an Fe-W or Fe-W-Mo alloy. Slag is produced upon the addition of lime and clay. This slag extracts nearly all of the Thorium contained in the metal melt. Selective dissolution of Tungsten in aqueous alkaline medium after oxidation of the metal to the hexavalent state by heating the scrap in air at temperatures of 500°C to 600°C. The resulting oxides are treated with sodium hydroxide solution. Tungsten and Molybdenum oxides are readily dissolved, while Thorium oxide together with silicon and aluminum compounds remain insoluble and are separated by filtration. Sodium tungstate solution is further processed by the usual hydrometallurgical tungsten mill process. Oxidation and dissolution of Tungsten can be achieved in one step by an electrochemical process. Thus, thoriated Tungsten scrap is used as an anode in an electrolysis cell, while sodium hydroxide or ammonia serve as electrolyte. After dissolution of Tungsten, the solids are separated from the liquid by filtration. With the electrochemical process, treatment of Tungsten-Thoria scrap can be achieved with high throughput in rather small reactors at moderate temperatures and ordinary pressure. The Tungsten solution exhibits high purity. Another process which we examined in detail is the separation of radium from rare earth compounds. Radium was separated by co-precipitation with barium sulfate from rare earth chloride solutions. The efficiency of the separation is strongly pH-dependent. Again, the valuable rare earth compound can be reused, and the radioactive elements are concentrated.

In recent years, various fuel types or targets, free from U-238, have been proposed for the transmutation of plutonium and the minor actinides. Yttrium-stabilised zirconia is currently under investigation because of its small neutron cross section and is high solubility with PuO 2 and AmO 2 . The aim of the present work is to investigate the behaviour of Yttrium Stabilised Zirconium Oxide in presence of Cerium, as simulate of minor actinides. Several oxide materials composed of Zr, Y, Ce (5–25%), have been synthesised by coprecipitation method at different calcination temperature and the products have been characterised by Optical Microscopy (OM), Thermogravimetry (TG) coupled with Thermal Differential Analysis (DTA), X-ray Line Diffraction (XRD), Scanning Electron Microscopy (SEM) Energy Dispersive X-Ray analyser (EDX) etc. The investigations show that, the structure of the powder, characterised by the phase condition, the lattice parameter, the crystalline size and the lattice distorsion, varies with the CeO 2 content, and has an influence on the sinterability. “Mild” methods of powder preparation (drying, low-temperature calcination, wet grinding by attrition) were used before sintering. For all investigated Cerium concentrations, the optimal parameter of pressing and sintering were determined. For a wide Cerium concentration region, from 5 to 20 wt%, the density of compact materials is 92–95% of TD, the microhardness 13.5 GPa and the pellets have a good morphology, grains and pores being homogeneous distributed in material.