In France, two options can be considered to handle the Very Low Level Waste (VLLW) and the Low Level Waste (LLW). The first one is the incineration at CENTRACO facility and the second one is the disposal at ANDRA sites. The waste acceptance in these two channels is dependent upon the adequacy between the waste characteristics (physical chemistry and radiological) and the channel specifications. If the waste characteristics and the channel specifications (presence of significant quantities of halogens, complexants agents, organic components… or/and high activity limits) are incompatible, an alternative solution have to be identify. It consists of a waste pre-treatment process. For Cadarache LOR (Liquides Organiques Radioactifs) waste streams, two radioactive scintillation cocktails have to be treated. They are composed of a mix of organic liquids and water: for the first one, 19 % of organic compounds (xylene, mesitylene, diphenyloxazole, TBP…) and 86.9 % of water, and for the second one, 23 % of organic compounds (TBP…) and 77 % of water. They contain halogens (chlorine and fluorine), complexants agents (nitrate, sulphate, oxalate and formate) and have got αβγ spectra with mass activities equal to some 100 Bq/g. Therefore, tritium is also present. As a consequence, in order for storage acceptance at the ANDRA site, it is necessary to pre-treat the waste. An adequate solution seems to be a solidification process using NOCHAR polymers. Indeed, NOCHAR polymers correspond to an important variety of products applied to the treatment of radioactive aqueous and organic liquids (solvent, oil, solvent/oil mixing …) and sludge through a mechanical and chemical solidification process. For Cadarache LOR, N910 and N960 respectively dedicated to the organic and aqueous liquids solidification are considered. With the N910, the organic waste solidification occurs in two steps. As the organic liquid travels moves through the polymer strands, the strands swell and immobilize the liquid. Then as the polymer-organic cure, over time, the polymer continues to collapse on the organic to create a permanent chemical bond. The N960 has the ability to absorb aqueous waste up to 100 times its own weight. It creates a strong mechanical bond which permanently traps the contamination imbedded in the aqueous liquids. Therefore, these two NOCHAR polymers seem to be able to constitute a suitable solidification matrix for a final acceptance in storage on ANDRA sites. In order to validate the solidification process using NOCHAR polymers as an acceptable solution for Cadarache LOR treatment, some solidification tests realised with N910 and N960, have been carried out for different Waste/Polymer ratios. To determine the best Waste/Polymer ratio and the optimal experimental parameters, exudation tests have been made. Indeed, the process prevents leaching and it results in the absence residual free organic or aqueous liquid which is forbidden in storage by ANDRA specifications. With these results, the obtained scientific data constitute a fundamental basis of an ANDRA agreement. As a conclusion, the aim of this study is to demonstrate that the pre-treatment by solidification using NOCHAR polymers can constitute a solution for Cadarache LOR handling and more generally, for various organic and mixed organic/aqueous waste which can not be directly acceptable at CENTRACO facility or at ANDRA storage sites. So then, this study is a solid background to demonstrate the feasibility of the waste pre-treatment by solidification with NOCHAR polymer and to encourage this process development.

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.

Ongoing experimental work has been underway at selected nuclear sites in the Russian State Atomic Energy Corporation (ROSATOM) during the past two years to determine the effectiveness, reliability, application and acceptability of high technology polymers for liquid radioactive waste solidification. The long term project is funded by the U.S. Department of Energy’s Initiatives for Proliferation Prevention (IPP) program. IPP was established in 1994 as a non-proliferation program of DOE / National Nuclear Security Administration and receives its funding each year through Congressional appropriation. The objectives of IPP are: • To engage former Soviet nuclear weapons scientists, engineers and technicians, currently or formerly involved with weapons of mass destruction, in peaceful and sustainable commercial activities. • To identify non-military, commercial applications for former Soviet institute technologies through cooperative projects among former Soviet weapons scientists, U.S. national laboratories and U.S. industry. • To create new technology sources and to provide business opportunities for U.S. companies, while offering commercial opportunities and meaningful employment for former weapons scientists. Argonne National Laboratory provides management oversight for this project. More than 60 former weapons scientists are engaged in this project. With the project moving toward its conclusion in 2012, the emphasis is now on expanding the experimental work to include the sub-sites of Seversk (SCC), Zheleznogorsk (MCC) located in Siberia and Gatchyna (KRI) and applying the polymer technology to actual problematic waste streams as well as to evaluate the prospects for new applications, beyond their current use in the nuclear waste treatment field. Work to date includes over the solidification of over 80 waste streams for the purpose of evaluating all aspects of the polymer’s effectiveness with LLW and ILW complex waste. Waste stream compositions include oil, aqueous, acidic and basic solutions with heavy metals, oil sludge, spent extractants, decontamination solutions, salt sludge, TBP and other complex waste streams. Extensive irradiation evaluation (up to 270 million rad), stability and leach studies, evaporation and absorption capacity tests and gas generation experimentation on tri-butyl phosphate (TBP) waste have been examined. The extensive evaluation of the polymer technology by the lead group, V.G. Khlopin Radium Institute, has resulted in significant discussion about its possible use within the ROSATOM network. At present the focus of work is with its application to legacy LLW and ILW waste streams that exist in a variety of sectors that include power plants, research institutes, weapons sites, submarine decommissioning and many others. As is the case in most countries, new waste treatment technologies first must be verified by the waste generator, and secondly, approved for use by the government regulators responsible for final storage. The polymer technology is the first foreign sorbent product to enter Russia for radioactive waste treatment so it must receive ROSATOM certification by undergoing irradiation, fire / safety and health / safety testing. Experimental work to date has validated the effectiveness of the polymer technology and today the project team is evaluating criteria for final acceptance of the waste form by ROSATOM. The paper will illustrate results of the various experiments that include irradiation of actual solidified samples, gas generation of irradiated samples, chemical stability (cesium leach rate) and thermal stability, oil and aqueous waste stream solidification examples, and volume reduction test data that will determine cost benefits to the waste generator. Throughout the course of this work, it is apparent that the polymer technology is selective in nature; however, it can have broad applicability to problematic waste streams. One such application is the separation and selective recovery of trans-plutonium elements and rare earth elements from standard solutions. Another application is the use of polymers at sites where radioactive liquids are accidently emitted from operations, thus causing the risk of environmental contamination.

In this study, the core-shell composite polymer for decontamination from the surface contamination was synthesized by the method of emulsion polymerization and blends of polymers. The strippable polymer emulsion is composed of the poly(styrene-ethyl acrylate) [poly(St-EA)] composite polymer, poly(vinyl alcohol) (PVA) and polyvinylpyrrolidone (PVP). The morphology of the poly(St-EA) composite emulsion particle was core-shell structure, with polystyrene (PS) as the core and poly(ethyl acrylate) (PEA) as the shell. Core-shell polymers of styrene (St)/ethyl acrylate (EA) pair were prepared by sequential emulsion polymerization in the presence of sodium dodecyl sulfate (SDS) as an emulsifier using ammonium persulfate (APS) as an initiator. Related tests and analysis confirmed the success in synthesis of composite polymer. The products are characterized by FT-IR spectroscopy, TGA that were used, respectively, to show the structure, the thermal stability of the prepared polymer. Two-phase particles with a core-shell structure were obtained in experiments where the estimated glass transition temperature and the morphologies of emulsion particles. Decontamination factors of the strippable polymeric emulsion were evaluated with the polymer blend contents.

Fine crystalline powders of KCuFC were immobilized with alginate gel polymers by sol-gel methods. The uptake properties of KCuFC-microcapsules (KCuFC-MC) were examined by batch and column methods. The size of KCuFC-MC particle was estimated to be about 1 mm in diameter, and KCuFC powders were uniformly dispersed in KCuFC-MC particles. The uptake rate of Pd 2+ for KCuFC-MC was attained within 3 d, and the uptake of Pd 2+ was found to be independent of the temperature and coexisting HNO 3 concentration. As for the breakthrough properties of Pd 2+ through a column packed with KCuFC-MC, a breakpoint of 5% breakthrough was enhanced with lowering of flow rate and independent of coexisting HNO 3 concentration. The Pd 2+ ions were selectively adsorbed in the KCuFC crystal phase, while other metal ions such as Ru(NO) 3+ and ZrO 2+ were absorbed in the alginate phase. High uptake percentage of 98.6% was obtained by using the dissolved solutions of spent fuel from FBR-JOYO (119 GWd/t, JAEA). The alginate film enclosing KZnFC was further prepared by using the support of cellulose filter paper, where the Pd 2+ ions were selectively adsorbed on the KZnFC-MC film. The alginate film enclosing insoluble ferrocyanides are predicted for the selective separation of Pd 2+ as an ion-exchange filter. Thus, the microcapsules enclosing insoluble ferrocyanides are effective for the selective separation of Pd 2+ from high-level liquid waste (HLLW).

Large amounts of liquid radioactive waste have existed in the U.S. and Russia since the 1950’s as a result of the Cold War. Comprehensive action to treat and dispose of waste products has been lacking due to insufficient funding, ineffective technologies or no proven technologies, low priority by governments among others. Today the U.S. and Russian governments seek new, more reliable methods to treat liquid waste, in particular the legacy waste streams. A primary objective of waste generators and regulators is to find economical and proven technologies that can provide long-term stability for repository storage. In 2001, the V.G. Khlopin Radium Institute (Khlopin), St. Petersburg, Russia, and Pacific Nuclear Solutions (PNS), Indianapolis, Indiana, began extensive research and test programs to determine the validity of polymer technology for the absorption and immobilization of standard and complex waste streams. Over 60 liquid compositions have been tested including extensive irradiation tests to verify polymer stability and possible degradation. With conclusive scientific evidence of the polymer’s effectiveness in treating liquid waste, both parties have decided to enter the Russian market and offer the solidification technology to nuclear sites for waste treatment and disposal. In conjunction with these efforts, the U.S. Department of Energy (DOE) will join Khlopin and PNS to explore opportunities for direct application of the polymers at predetermined sites and to conduct research for new product development. Under DOE’s “Initiatives for Proliferation Prevention” (IPP) program, funding will be provided to the Russian participants over a three year period to implement the program plan. This paper will present updated details of U.S. DOE’s IPP program, the project structure and its objectives both short and long-term, polymer tests and plications for LLW, ILW and HLW, and new product development initiatives.

The UK Nuclear Decommissioning Authority (NDA) is responsible for safe and secure management of spent nuclear fuel. Magnox spent fuel is held at some Magnox reactor sites and at Sellafield where it is reprocessed using a number of facilities. It is intended that all Magnox fuel will be reprocessed, as described in the published Magnox Operating Plan (MOP) [1]. In the event, however, that a failure occurs within the reprocessing plant, the NDA has initiated a programme of activities to explore alternative contingency options for the management of wetted Magnox spent fuel. Magnox fuel comprises metallic uranium bar clad in a magnesium alloy, both of which corrode if exposed to oxygen or water. Consequently, contingency options are required to consider how best to manage the issues associated with the reactivity of the metals. Questions of whether Magnox spent fuel needs to be dried, how it might be conditioned, how it might be packaged, and held in temporary storage until a disposal facility becomes available, all require attention. A review of potential contingency options for Magnox fuel was conducted by Galson Sciences Ltd, UKAEA and the NDA. During storage in the presence of water, the corrosion of Magnox fuel produces hydrogen (H 2 ) gas, which requires careful management. When uranium reacts with hydrogen in a reducing environment, the formation of uranium hydride (UH 3 ) may occur, which under some circumstances can be pyrophoric, and might create hazards which may affect subsequent retrieval and/or repackaging (e.g. for disposal). Other factors that may affect the choice of a viable contingency option include criticality safety, environmental impacts, security and Safeguards and economic considerations. At post-irradiation examination (PIE) facilities in the UK, Magnox spent fuel is dried as a result of storage in air at ambient temperatures. Early French UNGG (Uranium Naturel Graphite Gaz) fuel was retrieved from pond storage at Cadarache, dried using a hot gas drying technique, oxidised and packaged in sealed canisters and placed in interim storage at the CASCAD (CASemate CADarache) facility. In the US, spent fuels including the Zircaloy clad Hanford N-Reactor fuels were cold vacuum dried and Idaho legacy aluminium clad metallic uranium fuels were hot vacuum dried; the dried fuel was then packaged in sealed and vented canisters (at Hanford and Idaho, respectively) for interim storage. With regard to conditioning and packaging, several different approaches have been reviewed, including encapsulation in cementitious grout or polymer, high-temperature vitrification or ceramicisation, and solution in acid or alkali solution followed by cementation or vitrification (without reprocessing). All of these approaches require further research in order to be evaluated and developed further for application to formerly wetted Magnox fuel. A variety of containers have been developed for the transport, storage and/or disposal of spent fuel in radioactive waste management programmes worldwide. Wetted Magnox spent fuel could be packaged in a container, with reservations about the potential formation of UH 3 in a sealed environment where reducing conditions may develop. The applicability of different combinations of drying, conditioning and packaging techniques to the preparation of Magnox spent fuel for long-term storage and eventual disposal are discussed.

The presence of iron (iron oxide from carbon steel piping) buildup in Boiling Water Reactor (BWR) circuits and wastewaters is decades old. In, perhaps the last decade, the advent of precoatless filters for condensate blow down has compounded this problem due to the lack of a solid substrate (e.g., powdex resin pre-coat) to help drop the iron out of solution. The presence and buildup of this iron in condensate phase separators (CPS) further confounds the problem when the tank is decanted back to the plant. Iron carryover here is unavoidable without further treatment steps. The form of iron in these tanks, which partially settles and is pumped to a de-waterable high integrity container (HIC), is particularly difficult and time consuming to dewater (low shear strength, high water content). The addition upstream from the condensate phase separator (CPS) of chemicals, such as polymers, to carry out the iron, only produces an iron form even more difficult to filter and dewater (even less shear strength, higher water content, and a gel/slime consistency). Typical, untreated colloidal material contains both sub-micron particles up to, let’s say 100 micron. It is believed that the sub-micron particles penetrate filters, or sheet filters, thus plugging the pores for what should have been the successful filtration of the larger micron particles. Like BWR iron wastewaters, fuel pools/storage basins (especially in the decon. phase) often contain colloids which make clarity and the resulting visibility nearly impossible. Likewise, miscellaneous, often high conductivity, wastesteams at various plants contain such colloids, iron, salts (sometimes seawater intrusion and referred to as Salt Water Collection Tanks), dirt/clay, surfactants, waxes, chelants, etc. Such wastestreams are not ideally suited for standard dead-end (cartridges) or cross-flow filtration (UF/RO) followed even by demineralizers. Filter and bed plugging are almost assured. The key to solving these dilemmas is 1) to break the colloid (i.e., break the outer radius repulsive charges of the similar charged colloidal particles), 2) allow these particles to now flocculate (floc), and 3) form a type of floc that is more readily filterable, and, thus, dewaterable. This task has been carried out with the innovative application of electronically seeding the feed stream with the metal of choice, and without the addition of chemicals common to ferri-floccing, or polymer addition. This patent-pending new system and technique is called Seeding And Filtration Electronically, or the SAFE™ System. Once the colloid has been broken and flocking has begun, removal of the resultant floc can be carried out by standard, backwashable (or, in simple cases, dead-end) filters; or simply in dewaterable HICs or liners. Such applications include low level radwaste (LLW) from both PWRs and BWRs, fuel pools, storage basins, salt water collection tanks, etc. For the removal of magnetic materials, such as some BWR irons, an ElectroMagnetic Filter (EMF) was developed to couple with the ElectroCoagulation (EC), (or metal-Floccing) Unit. In the advent that the wastestream primarily contains magnetic materials (e.g., boiler condensates and magnetite, and hemagnetite from BWRs), the material was simply filtered using the EMF. Bench-, pilot- and full-scale systems have been assembled and applied on actual plant waste samples quite successfully. The effects of initial feed pH and conductivity, as well as flocculation retention times was examined prior to applying the production equipment into the field. Since the initial studies (Denton, et al, EPRI, 2006), the ultimate success of field applications is now being demonstrated as the next development phase. For such portable field demonstrations and demand systems, a fully self enclosed (secondary containment) EC system was first developed and assembled in a modified B 25 Box (Floc-In-A-Box) and is being deployed to a number of NPP sites. Finally, a full-scale SAFE™ System has been deployed to Exelon’s Dresden NPP as a vault cleanup demand system. This is a 30 gpm EC system to convert vault solids/sludges to a form capable of being collected and dewatered in a High Integrity Container (HIC). This initial vault work will be on-going for approximately three months, before being moved to additional vaults. During the past year, additional refinements to the patent pending SAFE™ System have included the SAFER™ System (Scalant and Foulant Electronic Removal) for the removal by EC of silica, calcium and magnesium. This has proven to be an effective enabler for RO, NF and UF as a pretreatment system. Advantages here include smaller, more efficiently designed systems and allowed lower removal efficiencies with the removal of the limiting factor of scalants. Similarly, the SAFE™ System has been applied in the form of a BAC-UP™ System (Boric Acid Clean-Up) as an alternative to more complex RO or boric acid recycle systems. Lastly, samples were received from two different DOE sites for the removal of totally soluable, TDS, species (e.g., cesium, Cs, Sr, Tc, etc.). For these applications, an ion-specific seed (an element of the SMART™ System) was coupled with the Cs prior to EC and subsequent filtration and dewatering, for the effective removal of the cesium complex and the segregation of low level and high waste (LLW & HLW) streams.

Spent ion exchanger media are considered to be problematic waste that, in many cases, requires special approaches and precautions during its immobilization to meet the acceptance criteria for disposal. The waste acceptance criteria define, among others, the quality of waste forms for disposal, and therefore will sometimes define appropriate treatment options. The selection of treatment options for spent ion exchange materials must consider their physical and chemical characteristics. Basically, the main methods for the treatment of spent organic ion exchange materials, following to pretreatment methods are: - Direct immobilization, producing a stable end product by using Cement, Bitumen, Polymer or High Integrity Containers, - The destruction of the organic compounds by using Thermochemical processes or Oxidation to produce an inorganic intermediate product that may or may not be further conditioned for storage and/or disposal, - The complete removal of the resin inner structural water by a thermal process. After a thorough technical economical analysis, Tractebel Engineering selected the Resin Hot Compaction Process to be installed at Tihange Nuclear Power Plant. The Resin Hot Compaction Process is used to make dense homogenous organic blocks from a wide range of particulate waste. In this process spent resins are first dewatered and dried to remove the inner structural water content. The drying takes place in a drying vessel that holds the contents of two 200 L drums (Figure). In the oil heated drying and mixing unit, the resins are heated to the necessary process temperature for the hot pressing step and then placed into special metal drums, which are automatically lidded and immediately transferred to a high force compactor. After high force compaction the pellets are transferred to a measuring unit, where the dose rate, height and weight are automatically measured and recorded. A volume reduction factor of approximately up to four (depending on the type of resins) is achievable using hot compaction techniques. This paper will describe the application of the Resin Hot Compaction Process, at Tihange NPP.

Non-standard wastes — those defined as being both hazardous waste under the United Kingdom Hazardous Waste Regulations 2005 [1] and radioactive under the Radioactive Substances Act 1993 [2] — pose particular, unique challenges for radioactive waste management organizations [3]. Treatment and disposal routes for such wastes are limited, in some cases non existent, and generally not cost effective. A non-standard waste of particular concern in the United Kingdom, and indeed on the Sellafield site, is that of radiologically contaminated waste oil. The optioning process for treatment of bulk contaminated waste oil on the Sellafield site has assessed a range of options including incineration, chemical decontamination, physical decontamination and immobilization. Immobilization has proved to be a potentially useful option for oil wastestreams that fail to meet waste acceptance criteria for incineration facilities. Experimental development work has been undertaken at Sellafield during 2006 to test the suitability of an innovative technology for the solidification of waste oil with a cross section of wastestreams from the site. These trials have demonstrated that this polymer system is able to successfully immobilize a range of aged, chemically and physically diverse contaminated oil wastestreams and thus provide a potential solution to the disposal problem posed by this wastestream.

For the conditioning of spent resins contaminated with radionuclides, such as: 137 Cs, 134 Cs, 60 Co, 58 Co, 57 Co, 54 Mn, etc., techniques of direct immobilization in cement, bitumen and organic polymers have been tested. The selected process was the bituminization using industrial bitumen, I 60–70, made in Romania, which had very good immobilizing characteristics. The paper presents stages of the research project, technical conditions for the process and advantages of the bituminization of spent resins.