Although cementation is a widely recognized solidification/stabilization process for immobilisation of Intermediate Level Radioactive Waste (ILRW), the low resistance to hyperalkaline pore waters compromises the effectiveness of the process when Portland Cement (PC) is employed. Moreover the manufacture of PC is responsible for significant CO 2 emissions. In this context, low pH cements are environmentally more suitable and have emerged as a potential alternative for obtaining secure waste forms. This paper summarises the achievements on development of low-pH cements and the challenges of using these new materials for the ILRW immobilisation. The performance of waste forms is also discussed in terms of radionuclides release. Reactive magnesium oxide and magnesium phosphate cements are emphasised as they feature important advantages such as consumption of available constituents for controlling acid-base reactions, reduced permeability and higher density. Additionally, in order to identify new opportunities for study, the long-term modelling approach is also briefly discussed.

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

Concentration of complexing acids solutions (oxalic acid, tartaric acid and citric acid) used in radioactive decontamination by chemical method affects both the mechanical stability and the chemical stability of cement matrix. The paper presents the works performed related to the chemical pretreatment of these organic acids solutions using as neutralizing agent Ca(OH) 2 . In this way it was possible to increase the concentration of organic acids solutions used and the soluble complex radionuclides passing in chemical precipitates, these affecting in a smaller manner the mechanical stability of the cement matrix. The chemical pretreatment the effluents improve the precipitation and conditioning performances by cementation. Were prepared compositions with complexing agents and compositions for oxidative degradation tests to simulate the concentrations of secondary radioactive waste obtained from the primary circuit decontamination of the VVR-S research reactor. It has been studied the influence of chemical pretreatment of complexing acids solutions of different concentrations on the setting time. Also it was determined the compressive strength of mortar samples in which were embedded these solutions of chemically pretreated organic acids. The results shown that an optimum cement–solution ratio doesn’t have a significant impact on the setting time or on the mechanical properties.

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.

A cement solidification process for treating sodium borate waste produced in pressurized water reactor (PWR) plants was studied. To obtain high volume reduction and high mechanical strength of the waste, simulated concentrated borate liquid waste with a sodium / boron (Na/B) mole ratio of 0.27 was dehydrated and powdered by using a wiped film evaporator. To investigate the effect of the Na/B mole ratio on the solidification process, a sodium tetraborate decahydrate reagent with a Na/B mole ratio of 0.5 was also used. Ordinary portland cement (OPC) and some additives were used for the solidificaiton. Solidified cement prepared from powdered waste with a Na/B mole ratio 0.24 and having a high sillica sand content (silica sand/cement>2) showed to improved uniaxial compressive strength.

The NDA’s Upstream Optioneering project has examined the potential implications of using an alternative type of waste package and its influence on the waste management lifecycle across the NDA estate. Robust Shielded Containers (RSCs) are radioactive waste packages that provide integral radiological shielding, reducing the need for remote handling. The robustness of the container could remove the need to immobilise waste by cement encapsulation. RSCs are routinely used to package ILW for interim storage in Germany and have fairly recently been considered for use in the UK because RSCs have the potential to simplify the waste management lifecycle and enable risk and hazard reduction. In particular, the current baseline (included in Magnox Limited lifetime plans) follows the strategy set out in the Magnox Operating Decommissioning Programme (MODP), in which Type II MOSAIK and Type VI Ductile Cast Iron Containers (DCICs) are used to package (in non-encapsulated form) most Magnox ILW arising prior to 2050. By considering representative types of Intermediate Level Waste (ILW) and exploring potential packaging options for these wastes, this paper identifies the factors that could differentiate between cases in which RSCs would, or would not, be an appropriate option. The potential role of RSCs across the waste management lifecycle is examined, from retrieval of waste through to emplacement at a Geological Disposal Facility (GDF), including consideration of other potential uses of RSCs, such as temporary storage of raw wastes for which appropriate treatment and conditioning measures have yet to be developed.

In large cement-based structures such as a near surface disposal facility for radioactive waste voids and cracks are inevitable. However, the pattern and nature of cracks are very difficult to predict reliably. Cracks facilitate preferential water flow through the facility because their saturated hydraulic conductivity is generally higher than the conductivity of the cementitious matrix. Moreover, sorption within the crack is expected to be lower than in the matrix and hence cracks in engineered barriers can act as a bypass for radionuclides. Consequently, understanding the effects of crack characteristics on contaminant fluxes from the facility is of utmost importance in a safety assessment. In this paper we numerically studied radionuclide leaching from a crack-containing cementitious containment system. First, the effect of cracks on radionuclide fluxes is assessed for a single repository component which contains a radionuclide source (i.e. conditioned radwaste). These analyses reveal the influence of cracks on radionuclide release from the source. The second set of calculations deals with the safety assessment results for the planned near-surface disposal facility for low-level radioactive waste in Dessel (Belgium); our focus is on the analysis of total system behaviour in regards to release of radionuclide fluxes from the facility. Simulation results are interpreted through a complementary safety indicator (radiotoxicity flux). We discuss the possible consequences from different scenarios of cracks and voids.

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.

Because Ca-dissolution from cementitious material is considered a source of long-term alteration of the performance of radioactive waste repositories, much research including dissolution tests has been conducted on this topic. These studies have introduced models such as Atkinson’s model [1] to calculate the leaching of cementitious material. These models have been used to verify that the results of many studies do represent the alteration of cementitious minerals and the composition of the leachate. They have also been used to make numerous estimates of long-term mineralogical alteration in repositories, such as cement-clay interaction in cementitious barrier systems, and to evaluate the change in repository performance. However, immersion tests using bulky cementitious material have often indicated that the actual alteration of cementitious material might be slower than the rates calculated by these models. This difference may be due to a change of mass-transport characteristics, either in the bulky cementitious material or at the interfaces with other materials. In this study, a mineralogical analysis was conducted on two types of old concrete. Drilled cores from the foundations of rotary kilns at two cement factories were collected beneath the groundwater level. Both concrete structures were made from Japanese ordinary Portland cement (OPC), which is similar to European type 1 cement. One structure had been set into a fresh groundwater environment for 78 years (78-F), and the other had set sunk into a saline environment for 98 years (98-S).

A “realistic alteration model” is needed for various cementitious materials. Hypothetical settings of mineral composition calculated based on the chemical composition of cement, such as Atkins’s model, have been used to estimate the alteration of cementitious material. However, model estimates for the concentration of certain elements such as Al and S in leachate have been different from experimental values. In a previous study, we created settings for a mineralogical alteration model by taking the initial chemical composition of cementitious materials from analysis results in experiments and applying their ratios to certain hydrated cement minerals, then added settings for secondary generated minerals in order to account for Ca leaching. This study of alteration estimates for ordinary portland cement (OPC) in groundwater showed that the change in Al and S concentrations in simulated leachate approached values for actual leachate[1]. In the present study, we develop an appropriate mineral alteration model for blended cementitious materials and conduct batch-type leaching experiments that use crushed samples of blast furnace slag cement (BFSC), silica cement (SC), and fly ash cement (FAC). The cement blends in these experiments used OPC blended with blast furnace slag of 70 wt.%, silica cement consisting of an amorphous silica fine powder of 20 wt.%, and fly ash of 30 wt.%. De-ionized water was used as the leaching solution. The solid-liquid ratios in the leaching tests were varied in order to simulate the alteration process of cement hydrates. The compositions of leachate and minerals obtained from leaching tests were compared with those obtained from models using hypothetical settings of mineral composition. We also consider an alteration model that corresponds to the diversity of these materials. As a result of applying the conventional OPC model to blended cementitious materials, the estimated Al concentration in the aqueous solution was significantly different from the measured concentration. We therefore propose an improved model that takes better account of Al behavior by using a more reliable initial mineral model for Al concentration in the solution.

A method, in which incinerated ash is solidified with a cement material, has been developed to dispose radioactive incinerated ash waste. In order to bury the solidified product, it is required that elution of hazardous heavy metals included in the ash from the solidified products is inhibited. In this study, the elution behavior of the heavy metals from the synthetic solidified products, which included Pb(II), Cd(II), and Cr(VI) and were prepared using ordinary portland cement (OPC), blast furnace slag cement (BFS), or a cement material that showed low alkalinity (LA-Cement), was investigated. Several chemicals and materials were added as additive agents to prevent the elution of the heavy metals. When OPC was used, Cd elution was inhibited, but Pb and Cr were not enough even using the additive agent examined. FeSO 4 and Na 2 S additive agents worked effective to inhibit elution of Cr. When BFS was used, the elution of Pb, Cd and Cr was inhibited for the all products prepared. In the case of LA-Cement, the elution of Pb and Cd was inhibited for the all products, but only the product that was added FeSO 4 showed good result of the elution of Cr.

Traditional methods of processing evaporator concentrates from Nuclear Power Plants are evaporation and cementation. These methods allow transforming a liquid radioactive waste into a more inert form, suitable for a final disposal. To assess the safety for disposal of radioactive mortar-waste composition, the leaching of 137 Cs from immobilized radioactive evaporator concentrate into a surrounding fluid has been studied. Leaching tests were carried out in accordance with a method recommended by IAEA. Curing conditions and curing time prior to commencing the leaching test are critically important in leach studies since the extent of hydration of the cement materials determines how much hydration product develops and whether it is available to block the pore network, thereby reducing leaching. Incremental leaching rates R n (cm/d) of 137 Cs from evaporator concentrates after 1825 days were measured. The results presented in this paper are examples of results obtained in a 30-year concrete testing project which will influence the design of the engineer trenches system for future central Serbian radioactive waste storing center.

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.

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.

In Japan, any high level radioactive waste (HLW) repository is to be constructed at over 300 m depth below surface. Tunnel support is used for safety during the construction and operation, and shotcrete and concrete lining are used as the tunnel support. Concrete is a composite material comprised of aggregate, cement and various admixtures. Low alkaline cement has been developed for the long term stability of the barrier systems whose performance could be negatively affected by highly alkaline conditions arising due to cement used in a repository. Japan Atomic Energy Agency (JAEA) has developed a low alkaline cement, named as HFSC (Highly Fly-ash Contained Silicafume Cement), containing over 60 wt% of silica-fume (SF) and fly-ash (FA). HFSC was used experimentally as the shotcrete material in construction of part of the 140m deep gallery in the Horonobe Underground Research Laboratory (URL). The objective of this experiment was to assess the performance of HFSC shotcrete in terms of mechanics, workability, durability, and so on. HFSC used in this experiment is composed of 40 wt% OPC (Ordinary Portland Cement), 20 wt% SF, and 40 wt% FA. This composition was determined based on mechanical testing of various mixes of the above components. Because of the low OPC content, the strength of HFSC tends to be lower than that of OPC. The total length of tunnel using HFSC shotcrete is about 73 m and about 500 m 3 of HFSC was used. The workability of HFSC shotcrete was confirmed in this experimental construction.

In a laboratory study, 15% (w/w) of ordinary portland cement (OPC), black portland cement (BPC) and puculanic cement (PC) combined with various cement additives were used for solidification/stabilization (S/S) of Cd, Pb, Zn, Cu, Ni and As contaminated soils from the former industrial site. Soil formed solid monoliths with all cements and cement additives. S/S effectiveness was assessed by measuring the mechanical strength of the monoliths, concentrations of metals in deionised water and TCLP (Toxicity Characteristic Leaching Procedure) soil extracts, and mass transfer of metals. Concentrations of Cd, Pb, Zn and Ni in water extracts from S/S soils generally decreased, while concentrations of As and Cu increased. Concentrations of Cd, Pb, Zn, Cu and Ni in the TCLP extracts from S/S mixes were lower than from original soil, while the extractability of As from S/S mixes increased. Overall, the concentration of metals in TCLP solution, obtained after extraction of the S/S mixes, was below the regulatory limits. S/S greatly reduced the mass transfer of Cd (up to 21.5-times), Pb (up to 114-times) and Zn (up to 999-times). Mass transfer of Ni was generally also reduced, while that of Cu and As increased in some S/S mixes. In general, monoliths made with OPC present higher cumulative mass of PTMs leached than those with BPC and PC.

Bentonite is one of the safety-critical components of the engineered barrier system for the disposal concepts developed for many types of radioactive waste. However, bentonite — especially the swelling clay component that contributes to its essential barrier functions — is unstable at high pH. To date, results from laboratory tests on bentonite degradation have been ambiguous as the reaction rates are so slow as to be difficult to observe. As such, a key goal in this project is to examine the reaction of natural bentonites in contact with natural hyperalkaline groundwaters to determine if any long-term alteration of the bentonite occurs. Ophiolites have been identified as sources of hyperalkaline groundwaters that can be considered natural analogues of the leachates produced by some cementitious materials in repositories for radioactive waste. At the Zambales ophiolite in the Philippines, widespread active serpentinisation results in hyperalkaline groundwaters with measured pH values of up to 11.7, falling into the range typical of low-alkali cement porewaters. These cements are presently being developed worldwide to minimise the geochemical perturbations which are expected to result from the use of OPC-based concretes (see Kamei et al., this conference, for details). In particular, it is hoped that the lower pH of the low-alkali cement leachates will reduce, or even avoid entirely, the potential degradation of the bentonite buffer which is expected at the higher pH levels (12.5 and above) common to OPC-based concretes. During recent field campaigns at two sites in the Zambales ophiolite (Mangatarem and Bigbiga), samples of bentonite and the associated hyperalkaline groundwaters have been collected by drilling and trenching. At Mangatarem, qualitative data from a ‘fossil’ (i.e. no groundwater is currently present) reaction zone indicates some alteration of the bentonite to zeolite, serpentine and CSH phases. Preliminary reaction path modelling suggests that the zeolites could have been produced as a product of smectite reaction in the hyperalkaline groundwaters. Although not included in this calculation to date, the CSH phases identified are completely consistent with reaction of clays with hyperalkaline groundwaters, as seen at other sites worldwide. At the Bigbiga site, an active hyperalkaline groundwater/bentonite reaction zone (at the base of the bentonite deposit) has recently been identified and a drilling campaign is planned for late autumn 2010.

In many radioactive waste repository designs, cement-based materials are expected to dominate the repository and models of cement evolution predict that leaching of the cementitious material in the repository by groundwater will produce an initial stage of hyperalkaline (pH∼13.3) leachates, dominated by alkali hydroxides, followed by a longer period of portlandite and C-S-H (CaO-SiO 2 -H 2 O) buffered (pH∼12.5) leachates. It has also been predicted that, as the hyperalkaline porewater leaches out of the near-field, significant interaction with the repository host rock and bentonite buffer and backfill may occur. This could possibly lead to deterioration of those features for which the host rock formation and bentonite were originally chosen ( e.g. low groundwater flux, high radionuclide retardation capacity etc). The precise implications of cement leachate/repository host rock interaction has been studied in the laboratory and in underground research laboratories (URLs) and this work has been supported by study of natural cements in Jordan. These natural cements have been produced by the combustion of organic-rich clay biomicrites and are very close analogues of industrial cement. Following interaction with groundwaters, natural hyperalkaline leachates are produced and these move out of the cement into the surrounding host rock, subsequently interacting with and altering it.

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.