The Fukushima Dai-ichi and Dai-ni nuclear power station with 4 GE-BWRs units at one site and 2 BWR units respectively co-located on the north-central eastern coast of Japan withstood a 9.0 earthquake and a large-scale tsunami on March 11, 2011. All six units were constructed via a GE/Hitachi/Toshiba collaboration from 1967-1979. In spite of the immediate shut down of all units based on ground-level acceleration and decay heat cooling for some 30-45 minutes, loss-of-offsite-power by ingress of water into the diesel generators&’ pit, initiated loss-of-coolant accident; overall as a ‘beyond design basis accident&’. Further, all units faced unanticipated challenge of cooling spent fuel pools situated above the reactors in lightly-structured buildings. Several hydrogen explosions later and more than a year since ‘3/11&’, the utility (TEPCO) and the Japanese Government are now facing a 20- to 30-year cleanup effort. Evidence suggests that 3 reactor cores have partially-to-fully melted. The scale of the recovery, restoration and remediation effort will be very large. The accident has refocused attention on the need for an inherently safe LWR and ‘accident tolerant&’ fuel and cladding materials; that is, to mitigate the progression of a severe accident. However, in the near term at Fukushima, one of the challenges will be characterizing the state of the partially-to-fully melted (UO₂) fuel, including the state of the mixed oxide (UO₂ and PuO₂) fuel in Unit 4. The fuel is housed in Zircaloy cladding; the Zr reacts with steam at elevated temperatures to produce H₂. We anticipate that based on the state of ‘burn-up&’ of the fuel at the time of the accident and lack of decay heat cooling, the thermal condition of the fuel, cladding and coolant dictated the eventual state of fuel/clad/coolant. Since a detailed progression sequence will contain large uncertainties, the material forensics will be macroscopic and circumstantial in nature. However, microscopic information is needed. The presentation will provide a quick perspective on the Fukushima accident, technical and non-technical lessons learned and issues of interest to the nuclear materials community.

The damage to the Fukushima Dai-ichi nuclear power plant by the Great Tohoku earthquake and tsunami resulted in considerable contamination, both on- and off-site. Work is already advanced to implement regional decontamination, with a special focus on allowing the evacuated population to return and re-establish normal lifestyles as soon as possible. After decay of shorter-lived isotopes, the challenges for off-site remediation mainly involve radiocaesium isotopes (-134 & -137). As remediation on this scale and in such a geographic setting is unprecedented, JAEA was adopted by the Government to conduct decontamination pilot projects at model sites. These projects (1st covering 2 sites with lower contamination levels; 2nd including 16 sites in 11 municipalities, some with significantly higher contamination) allowed acquisition of technical data and knowledge and development of the integrated expertise required to support the planned regional decontamination. Despite tight boundary conditions in terms of timescale and resources, the decontamination pilot projects provide a good basis for developing recommendations on how to assure clean-up efficiency and reduce time, cost, subsequent waste management and environmental impact. This can be summarised in terms of: 1. Site characterization and data interpretation; Measurement approaches involved both tailoring of existing technology for Japanese conditions and development of new tools. Resultant radio-Cs maps and depth profiles were particularly useful to guide remediation planning. 2. Clean-up; Although the majority of the effort involves manual washing and contaminated material removal using conventional technology, methods that can improve decontamination while decreasing volumes of waste were successfully tested. A key challenge for sites with complex topography and land use was quantifying the extent of decontamination. 3. Waste handling and management; Waste was reduced in volume to the maximum extent possible - e.g. grinding / chipping of foliage. A number of different approaches were used for temporary storage of waste on the surface or in shallow pits. This paper discusses decontamination pilot projects achievements and their application for the next stage of Fukushima environmental restoration.

The earthquake, tsunami, and hydrogen explosions hitting the Fukushima Daiichi Nuclear Power Station left lots of radiation-contaminated debris on site from buildings, equipment, vehicles and so forth. Contaminated wastes such as effluent, co-precipitated sludge and spent filters are continuously produced from accident water treatment. Restoration works also generate contaminated wastes such as felled trees, dismantled debris, used anti-contamination clothing, many other forms of scraps, and dust every day after the accident. United workforce of members from the government, TEPCO, JAEA and other Japanese top authorities are now seriously planning, preparing and performing national level activities for cleanup of the station site. Many sorts of researches and developments for wastes&’ characterization and radioactive inventory estimation are ongoing. A variety of wastes are being sampled and analyzed for concentrations of alpha, beta, and gamma emitters. Safer and more effective sampling, transportation and analysis of wastes are under consideration. Much effort is also paid to implement effluent purification and safe release because of large volume of the stored effluent. Technological development required for mid-to-long-term on site storage of radioactive wastes, one of critically important milestones on the roadmap to cleanup, should become a very unique issue due to limitation of the storage area and air dose rate.

Bentonite is planned to be used in many countries as an important barrier in high-level waste repositories. Assessment of the barrier with regard to, inter alia, its ability to hinder transport of dissolved radionuclides leaking from a damaged canister containing spent nuclear fuel, requires quantitative data about the pore structure inside bentonite. The present NMR study was made in order to determine the number of distinguishable porewater phases in compacted water-saturated samples of MX-80 bentonite and Na-montmorillonite. The latter material was made by purifying MX-80 and thereafter saturating the pure material with Na ions. The samples were compacted to dry densities in the interval 0.7-1.6 g/cm3 and subsequently saturated with Milli-Q water or 0.1 M NaCl solution in equilibrium cells. The NMR measurements were performed with a high-field 270 MHz NMR spectrometer using a short inter-pulse CPMG method to study proton T1ρ relaxation. The measured relaxation curves were found to consist of one faster and one slower proton relaxation. Subsequent analysis of the data indicated that the faster relaxation was associated with interlayer (IL) water between montmorillonite unit layers, while the slower one was associated with non-interlayer (non-IL) water located outside the interlayer spaces. Our results show how the relative volumes of IL and non-IL water change as a function of the dry density of the investigated samples. Furthermore, the results show that in the case where the dry density of MX-80 equals that of Na montmorillonite, then the proportions of IL and non-IL water are different in the two materials. The reason for this is explained in terms of microstructural differences between the investigated samples.

Corrosion behavior is a key issue for the waste disposal of irradiated metals, such as the hulls and endpieces, and is considered to be a leaching souse of radionuclides including C-14. Under the disposal environment, with its anticorrosive condition, a little information about Zircaloy corrosion has been provided by use of the hydrogen measurement technique. [1] In the present study, long-term corrosion tests of Zircaloy 4 and Zircaloy 2 were performed in an estimated disposal condition (a dilute NaOH solution of pH = 12.5 at 303 K) using the gas flow system for 1500 days. The corrosion rate, which was determined by measuring gaseous hydrogen and the hydrogen absorbed into Zircaloys, decreased with immersion time and was lower than that used in the performance assessment of 2×10^-2 mu;m/y. [2] (1500-day values: 5.84×10^-3 and 5.66×10^-3 mu;m/y for Zircaloy 4, 750-day values: 9.93×10^-3 mu;m/y for Zircaloy 2) The difference in corrosion behavior between Zircaloy 4 and Zircaloy 2 was negligible. The average values of hydrogen absorption ratio for Zircaloy 4 and Zircaloy 2 during corrosion were 91% and 94%, respectively. The hydrogen generation kinetics both gas evolution and absorption into metal shows a parabolic curve (Zircaloy 4: the gaseous hydrogen per unit surface area (atomic mmol/m2) = 0.0034 × t^0.55 and the absorbed hydrogen per unit surface area (mmol/m2) = 0.04 × t^0.56; t is test time in day). This indicats that the diffusion process is controlling the Zricaloy corrosion in the present early corrosion stage, the oxide film in which is limited to approximately 25 nm thick and may therefore be in a form of dense tetragonal zirconia. The corrosion behavior will be also discussed with the C-14 leaching data from irradiated Zircaloy 4. [3] This research is a part of “Research and development of processing and disposal technique for TRU waste containing I-129 and C-14 (FY2011)” under a grant from the Japanese Ministry of Economy, Trade and Industry (METI). [1] T. Sakuragi et al., Corrosion Rates of Zircaloy 4 by Hydrogen Measurement under High pH, Low Oxygen and Low Temperature Conditions, Mater. Res. Soc. Symp. Proc. Vol. 1475 (2012). [2] FEPC and JAEA, Second Progress Report on Research and Development for TRU Waste Disposal in Japan (2007). [3] T. Yamaguchi et al., A Study on Chemical forms and Migration Behavior of Radionuclides in Hull Waste, Proc. Radioactive Waste Management and Environmental Remediation ASME, Nagoya, Japan (1999).

The safety assessment SR-Site is undertaken to assess the safety of a geologic repository of the KBS-3 type at the Forsmark site, Sweden. The assessment supports SKB&’s licence application for a final repository for spent nuclear fuel at Forsmark. If groundwater enters a damaged canister and comes in contact with the spent fuel, radionuclides may be released into the water. If the aqueous concentration of an element reaches saturation with respect to the solid phase, then its solubility limit is attained and the element will precipitate. As a result, only the aqueous fraction of the element may migrate with the water flowing from the canister while the fraction that has precipitated remains in the canister. The key factors that affect the elemental solubility limits were identified as: 1) the assumed solubility limiting phase, 2) the geochemical conditions inside the damaged canister and 3) the thermodynamic database used. Solubility limiting phases were selected by an “expert judgement”, favouring phases that would be likely to precipitate without any kinetic restrictions. The geochemical conditions inside the damaged canister were assumed to be identical to the conditions in the groundwater with the exception that the redox conditions were controlled by the magnetite/goethite equilibrium. The thermodynamic database used was the Nagra/PSI Chemical Thermodynamic Data. To produce probability density functions for elemental solubilities, the Simple Functions tool was developed. Simple Functions performs geochemical equilibrium calculations, but contains only the limited subset of data and reactions that is needed to calculate solubilities for the conditions that can be expected at the Forsmark site. Simple Functions was used in combination with the @risk software to fast and efficiently produce the solubility data. The assessment in SR-Site covers 6 000 canister positions and the assessment period is one million years. This means that there will be a natural spatial and temporal variability in the composition of the groundwater. To handle this, the solubility limits for the safety assessment were calculated with a set consisting of 25% of groundwater compositions representing the temperate climate, 25% representing the permafrost climate, 25% representing glacial climate and 25% representing submerged climate. For the uncertainties in thermodynamic data a normal distribution was applied to the equilibrium constants (mu; = log₁₀K⁰ and σ = (Δlog₁₀K⁰) / 2). The relative importance of variability in groundwater composition compared to uncertainty in thermodynamic data was evaluated by keeping either the groundwater composition or the thermodynamic data constant. The results showed that uncertainty in thermodynamic data has a bigger impact on the results for almost every radioelement. The sole exception to this is radium, which happens to be the most safety critical element, where variability in water composition has a somewhat larger impact.

In the French HLW management strategy, it is expected to store around 40,000 nuclear glass canisters arising from spent fuel reprocessing in a deep geological disposal using a multi-barrier concept: nuclear glass is poured into a stainless steel canister and the resulting system is placed in a low-alloy steel overpack, directly strored in a 100 m thick clayey host rock located 500 m below the surface. Consequently, source term resulting from interactions between the nuclear glass, the solution saturating the media and the near-field materials (iron, corrosion products, clay) must be assessed . In this study, glass - iron or corrosion products interactions were investigated in a clayey environment to better understand the mechanisms and driving forces controlling the glass alteration. Integrated experiments involving glass - metallic iron or magnetite - clay stacks were run at laboratory scale in anoxic conditions for two years. The interfaces were characterized by a multiscale approach using SEM-EDS, TEM, microRaman spectroscopy and STXM at the SLS Synchrotron. We specifically focused on the influence of the glass - iron source distance on the morphology and chemistry of glass alteration layers, and the valence state of iron in the different zones of the glass / iron source interface. Characterization of glass alteration patterns on cross sections revealed various morphologies or microstructures and an increase of the glass alteration with the proximity between the glass and the source of iron (iron or magnetite) due to the consumption of the silica coming from the glass alteration. In case of magnetite, the silica consumption is mainly driven by a sorption of silica onto the magnetite. However, some simulations using GRAAL [2] show that silica sorption on magnetite is not the only mechanism driving the glass alteration, Fe-silicates precipitation could also occur as it is shown by the alteration layer characterization. For experiments having metallic iron, the silica consumption seems to be strongly driven by silicates precipitation including Fe and Fe/Mg when the Fe is not enough available. Moreover, in addition to Fe-silicates observed at the surface of the gel layers, iron is incorporated within the gel probably as nanosized precipitates (Fe-oxyhydroxide or Fe-silicates) which could affect its transport properties.Those results highlighted the impact of the distance glass - iron source and the nature of the iron source which drive the process consuming the silica coming from the glass alteration. Such silica consumption, limited by the transport, does not allow the system to be saturated regarding the silica nor to form protective gel layer leading to higher glass alteration rate than without iron.The new data may imply some consequences when considering the long-term behavior of glass in geological disposal conditions.

Monazite (CePO4), apatite [Ca5(PO4)3(OH)], and other phosphate minerals are able to contain high concentrations of actinides, lanthanides, and other elements and isotopes (e.g., 90Sr) found in nuclear waste. Therefore, because of their stability and high sorptive capacity, phosphate minerals or phosphate-silicate solid solutions have been proposed as waste-forms, backfill, or overpack materials in nuclear waste repositories. We propose that bioapatite (dahllite), a form of carbonate-apatite found in bones and teeth of living vertebrates, be used as overpack or backfill material. Vertebrate bones are composed of approximately 70% bioapatite mineral and 30% organic collagen matrix. In bioapatite, CO32- substitutes for PO43-. The charge deficiency is usually compensated by omission of calcium or substitution of monovalent cations, producing a defect structure. Carbonate apatite crystals in bone are poorly crystalline, plate- or tablet-shaped and extremely small, with average dimensions of 50 x 25 x 2 to 4 nm and very large specific surface areas of ca. 240 m2/g. The large specific surface areas result from the small crystal size and high internal matrix surface areas and porosities of collagen-free bone. In apatite-containing backfill, concentrations nuclear waste species may be controlled either by solubility of their phosphate minerals or by sorption on apatite. Bioapatite is more soluble than hydroxy- or fluorapatite and is highly reactive. The rate of bioapatite dissolution is faster than that of fluorapatite or carbonate fluorapatite, and is constant at ca. 4.3 x 10-10 mol m-2 s-1 between pH 4 and 8 at 22°C. In contrast, the dissolution rate of sedimentary carbonate fluorapatite is slower and hydrogen ion-dependent (n ~ 0.6), decreasing until ca. pH 7. Therefore, rapid bioapatite dissolution, constant over a wide range of pH, and higher solubility would produce higher dissolved phosphate concentrations and lower near-field waste concentrations. In near-neutral pH solutions, measured sorption constants (Kd) between apatite and uranium and lanthanides range from ca. 5 x 105 to 2 x 106. Therefore, sorption could significantly decrease dissolved waste concentrations. Bioapatite may actively sorb and remove waste materials for long periods. Based on measured concentration gradients in marine and terrestrial fossils, periods of uranium and lanthanide incorporation have been calculated for bioapatite in fossils using Fick&’s second law. Diffusion and incorporation periods range between ca. 1 ka, in fully saturated, marine environments, to ca. 80 ka, in intermittently saturated terrestrial environments. Therefore, bioapatite may scavenge radioisotopes from solution over long periods. Adsorption and incorporation of fluoride and other trace elements and diagenetic growth of larger crystals decreases apatite solubility and reactivity and will allow wastes to be sequestered for millions of years.

The U.S. Department of Energy&’s Office of Nuclear Energy (DOE-NE) is addressing technical, cost, safety, security and regulatory issues through research, development, and demonstration activities to ensure that nuclear energy remains a compelling and viable energy option for the U.S. A significant aspect of this research is aimed at enabling sustainable nuclear fuel cycle options along with developing used nuclear fuel (UNF) management strategies and technologies to support the U.S. federal government&’s responsibility to manage and dispose of the U.S. commercial UNF and associated high level waste (HLW). This talk will describe the programmatic drivers, prioritization, and conduct of waste forms research as part of an integrated Fuel Cycle Research and Development (FCRD) program to enable options for managing the back end of the U.S. nuclear fuel cycle. As processes to recycle one or more UNF components are developed, unique waste streams arise that must be managed in a safe, environmentally friendly, and cost effective way. Although borosilicate glass is well demonstrated for the treatment of HLW streams, it still remains a potentially problematic process if not thoroughly developed and demonstrated for a unique waste stream as evidenced by the process difficulties experienced at the Rokkasho Reprocessing Plant (RRP). In addition, glass is not an optimal waste form for many of the steams being generated by potential separations technologies (e.g., undissolved solids and salts from electrochemical processing, gaseous fission product streams, and technetium). New waste forms must be developed, tested, and demonstrated for these streams concurrently with the separations technology development. The performance of waste forms is of paramount importance and opportunities exist to improve waste form / disposal site system. Uncertainties in the mechanisms dictating glass performance at long times has led to significant conservatism in current corrosion estimates. A recently initiated international collaboration will not only quickly advance the technical understanding of these processes, but an international consensus on corrosion rate will create the scientific confidence in less conservative models necessary to satisfy public and qualification requirements. Once the rate law has been developed and demonstrated on HLW glass, the methods and approaches will be applied to other waste forms requiring long service life such as the forms for I-129, Tc-99, and TRU.

Nuclear energy can be a part of the answer to the worldwide increase in energy needs, while limiting green-house gas emissions. The development of sustainable nuclear options requires the highest safety level, economic efficiency, but also efficient use of natural uranium resource, safe and socially accepted final waste management, proliferation resistance. Both reactors and fuel cycle options are concerned: a key-guideline seems to be the development of recycling options, taking advantage of the energetic content of spent fuel component, and minimizing final waste amount and long-term hazards while consuming proliferation-sensitive elements. France has been operating for decades recycling options for used fuel management. Recycling plutonium in light water reactors MOX fuels provides about 10% of the French electricity; plutonium-free ultimate waste are safely immobilized in long-lasting glass, which should be - according to the roadmap fixed by the French Act about radioactive waste management- disposed in clay from 2025. But this scheme appears to be improved, and to be completed: increasing important amounts of depleted uranium and, to a less but very significant extent, spent MOX fuels, are waiting for further valorization. Attractive features of fast neutron physics reactors present them as the best tool for that, opening the way for a drastic extension of uranium natural resource. And fast neutron systems seem able, in addition, to significantly decrease long-term potential radio toxicity of the residual waste by drastically decreasing the minor actinide content of final waste. CEA launched, more than one decade ago, a large research program to prepare such future systems, and explore the diverse suitable options . This program was an answer to the French Parliament&’s request (two consecutive Acts from 1991 calling for research to explore radioactive waste management options), and consistent with the “generation four” international forum. A comprehensive scientific report is to be issued by the end of this year, to present the results of the work performed and -as asked by the French act about nuclear waste management- assess the “industrial perspectives” of different options. In the mean time, the French government decided to launch a prototype of a generation 4 reactor.. The ASTRID prototype (a sodium-cooled 600 MWe fast neutron reactor, to be commissioned in the early 2020&’s) is currently under design, in a broad cooperation frame: coupled with advanced fuel cycle options, this prototype will provide the opportunity to go ahead towards fully sustainable nuclear systems, even more efficient, and safer. This presentation gives an overview of the current status of the research and main outcomes obtained at CEA on these topics.

Actinide chemistry is at the centre of key issues to be faced by nuclear energy. Indeed, in addition to an increased safety of the reactors themselves, the acceptance of the nuclear energy is still closely associated to our capability to reduce the lifetime of the nuclear waste, to manage them safely in a long term disposal and to propose options for a better use of the natural resources. This is compulsory to demonstrate that it can contribute safely and on a sustainable way to answer the international increase in energy needs. Actually, spent fuel reprocessing can help to reach these objectives. But this cannot be achieved only by optimizing industrial processes through engineering studies. It is of a primary importance to increase our fundamental knowledge in actinide sciences in order to meet the needs of the future fuel cycles in terms of safety, fabrication and performance of fuels, reprocessing and long term waste management. Among EURATOM Framework Program FP7-Fission projects, the Integrated Infrastructure Initiative ACTINET-I3 and the Collaborative Project ACSEPT work together to improve our knowledge in actinides chemistry in order to develop advanced separation processes, but also to increase our knowledge on actinide material chemistry and the chemistry of the actinides in the environment. By offering transnational access to the main European nuclear research facilities, ACTINET-I3 aims at increasing the knowledge in actinide sciences by gathering all the expertise available in nuclear research institutes or university in Europe and giving them the opportunity to come and work in hot-labs (ITU, CEA-Atalante, KIT-INEhellip;) or beamlines (ESFR, ANKA, PSI). Every six months, a call for proposals allows scientists to candidate for short stays (up to three months) at pooled facilities to perform a Joint Research Project in actinide chemistry. ACSEPT is focused on the development of advanced separation processes, both aqueous and pyrochemical. Head-end steps, fuel refabrication, solvent treatment, waste management are also taken into account. In aqueous process development, options have been developed for the DIAMEX, SANEX and GANEX strategies. In pyrometallurgy, studies on actinide back-extraction from aluminium and exhaustive electrolysis allowed the validation of two flowsheets developed from more then 10 years in Europe. In both projects, efforts have been made to increase collaborations, mutualise and homogenise procedures and share good practices. Training and education initiatives including seminars, workshops, brainstorming meeting but also student exchanges and support to post-doctorate fellowships was a key point for maintaining and increasing a high expertise level in actinide separation sciences in Europe. The paper will present the main achievements of these two key projects of FP7-EURATOM-Fission.

Meeting the future energy needs while mitigating the anticipated global climate change requires promoting low carbon energy systems, i.e. renewables and nuclear. However, whatever the energy mix selected, it will only develop if it meets the requirements of the sustainability, i.e. meeting simultaneously the social, economic and environmental criteria of viability, bearability and equitability. Each energy systems, among which nuclear, have therefore to be optimised regarding a set of criteria covering this three fields. Nowadays, most of the countries chose the so-called once-through cycle which basically considers spent nuclear fuel as a waste, whereas others like France, UK, Japan and soon China reprocess their spent fuel to recover the energetically-valuable material Pu (and partially U) to produce Mixed Oxide Fuel (MOX) to be irradiated in a second cycle (twice-through cycle). None of them are fully sustainable since they do not allow a complete use of the natural resource (thermal neutrons do not allow to efficiently use 238U), However, recycling U and Pu from spent fuel allows recovering 96% of the spent nuclear fuel which can be subsequently used in MOX and URE fuels to produce electricity: in France, 17% of natural uranium resource is hence yearly saved. Recycling actinides is also a significant contribution for the waste management issue. It allows both to specifically separate the sole ultimate waste (fission products and potentially minor actinides) and to condition them in a specific wasteform, the nuclear glass, which is designed to ensure the long-term confinement. It hence decreases significantly the waste volume (96% is recycled) and increases the long-term performance (the nuclear glass lifetime is ~1 million years). This paper aims to depict the relative environmental footprint of the two respective once-through and twice-through fuel cycles. Taking the French situation as an example, this paper will assess the respective figures of merits of both fuel cycles regarding the environmental impact, among which the waste management is the leading issue.

The concept Technology Readiness Levels was developed by NASA as a metric to support assessment of technology maturity and achieve consistent comparison of different technology types using a nine point scale: TRLs 1-3 relate to proof of basic science and feasibility, TRLs 4-6 relate to technology development and demonstration, and TRLs 6-9 relate to subsystem and system test and operation. In the UK the TRL system is applied by nuclear Site Licence Companies and other organisations to assess the maturity of decommissioning and waste clean up technologies. Vitrification technologies offer several advantages in this respect, compared to standard cementation, including: improved stability and passive safety of the conditioned product; substantially reduced product volume; scaleable deployment; and, potentially, lower whole life cycle costs. However, a perceived barrier to deployment of vitrification technologies for intermediate level waste treatment is a relatively low level of maturity. In this presentation, the feasibility of vitrifying UK intermediate level wastes and plutonium contaminated materials will be examined, in the context of the potential advantages highlighted above. Using selected case studies, the technology readiness of vitrification processes will be discussed, from examination of the design, prototyping and performance of glass compositions and demonstration using commercially available melting technologies at full scale using inactive simulants.

The development of suitable waste forms for waste produced by generation IV reactor designs is of critical concern for any future operations. Several glass compositions have been studied for their ability to encapsulate HTR fuels. The study focused on compositions known for their high aqueous durability. Encapsulation was achieved by cold press and sintering of glass powders mixed with HTR fuel. Compositional variations have been studied for their effect on aqueous durability, chemical compatibility, coating properties and mechanical properties. Sintering profiles capable of eliminating interconnected porosity have been developed. The aqueous durability of the sintered glasses has been shown to be comparable to that of precursor glasses and suitable for geological disposal. Mechanical properties of these sintered composites have been shown to be comparable or superior to those for currently employed HLW glasses. Sintering with a variety of glass compositions has been shown to have minimal negative chemical interactions when performed under a controlled atmosphere. This suggests sintered glass - HTR composites may provide a potential disposal route for spent HTR fuels. The glass composition has significant effects not only on aqueous durability, but also the coating properties of the final waste form to the HTR fuel and the matrix integrity. Compositional variations have been shown to have a marked effect on all aspects of product quality when used for encapsulation of HTR fuel and as such should be a focus for further work.

Durable waste forms for iodine-129, present in spent nuclear fuel, are being developed using several approaches. Safe disposal of iodine-129 is required due to its long half-life (>16 x 10⁶ years) and its harmful health effects. In spent nuclear fuel reprocessing schemes under development by the US DOE, iodine-129 vapor is passed over Ag-exchanged mordenite, a zeolite, to form AgI, which has low aqueous solubility. Because of the low melting point (558°C) and high vapor pressure at moderate temperatures of AgI, the maximum processing temperature for an AgI-containing waste is ~550°C. One type of waste form for AgI-mordenite that we have developed utilizes a low temperature sintering oxide glass powder that is mixed with ground AgI-mordenite, pressed into a compact and then sintered at 550°C to form a dense and durable waste form. Sintering as opposed to melting allows a more durable glass composition to be used. Aqueous leaching studies show a high degree of durability of this type of waste form, comparable to that of the borosilicate glasses commonly used in nuclear waste applications. We have also demonstrated the applicability of this approach to other wastes including pure AgI, AgI on titania nano-fibers and cesium-containing crystalline silico-titanates. In addition, we have developed processes to encase the waste form in a shell containing the same low-temperature sintering glass to further protect the environment. The shell can either be formed by dry pressing or, for a thinner shell, by tape casting. In either case, shell is sintered along with the AgI-mordenite containing core. To avoid CTE-mismatch cracking during cooling from the sintering temperature, amorphous silica powder is added to the glass comprising the shell. Mechanical testing data indicates that the shell&’s strength is comparable to that of the pure glass. We are also investigating the use of advanced I2 sorbent materials such as metal-organic framework materials (MOFs) that have a high iodine uptake capacity but are more temperature sensitive. For these materials (as well as AgI-mordenite materials) we have developed a room temperature process for forming a dense and durable waste form. In this approach, the iodine-containing material is simply mixed with an appropriate metal powder such as tin and then compacted at high pressure to yield a dense and robust waste form. If deemed necessary, this type of waste form could be encased in a tin canister for enhanced safety that is sealed by cold welding at room temperature. Sandia is a multiprogram laboratory operated by Sandia Corporation, a Lockheed Martin Company, for DOE's National Nuclear Security Administration under contract DE-AC04-94AL85000.

In the spent iodine filter which is generated from Japanese nuclear fuel reprocessing process, almost radioactive iodine (I-129) exists as silver iodide (AgI). The synthetic rock technique is a solidification treatment technique using hot isostatic press (HIP), in which the alumina adsorbent base material is synthesized to a dense solidified substance (synthetic rock), and I-129 is physically confined in the form of AgI in the alumina matrix. Thus, it is necessary to understand the matrix dissolution behavior to evaluate the iodine release behavior. Dissolution experiments of the matrix were carried out under various temperatures (35-80 degree C) and pHs (10-12.5) assumed in disposal condition. The test results showed that the dissolution rate of Al almost increases with temperature and pH. The dissolution rate constant was calculated from initial data when it was supposed that the dissolution of the matrix was a primary reaction. The natural logarithm of the rate constant showed a good linear correlation with pH and a reciprocal of absolute temperature. The ²⁷Al-NMR analysis was applied and it was shown that the main chemical species in those solutions was Al(OH)₄^-, indicating that the dissolution reaction of the matrix is described as Al₂O₃ + 2OH^- + 3H₂O → 2Al(OH)₄^-. From those results, the empirical equation of dissolution rate of the matrix as a function of the temperature and the pH was derived. The iodine release behavior from the synthetic rock will be evaluated in conjunction with the equation of dissolution rate of the matrix. This research is a part of “Research and development of processing and disposal technique for TRU waste containing I-129 and C-14 (FY2011)” under a grant from the Japanese Ministry of Economy, Trade and Industry (METI).

We have developed iodine-immobilized cement solidification process using the material of sulfate-added calcium aluminate cement (S-CAC). ¹²⁹I generated from reprocessing plant is processed to the chemical form of iodate ion, and fixed into oxyanion channels of ettringite (AFt ; (Ca₆[Al(OH)₆]₂ 24H₂O)(SO₄)₃ 2H₂O), which is one of major minerals formed in S-CAC material. In order to evaluate the iodine immobilization capability of the cement solid, continuously-dissolution accelerated test has been performed. The powder of the cement solid was repeatedly immersed with ion-exchanged water at a liquid-to-solid ratio (L/S) as accelerated dissolution tests simulating interaction with groundwater at the waste disposal site. The concentrations of iodine in the water measured the order of 10^-5 to 10^-3 mol/L along overall L/S. These concentration levels are significantly low compared to that in OPC (Ordinary Portland Cement) solid case, in which no confinement ability is expected. The solid phases were chemically analyzed at each L/S step to know alteration behavior of the mineral phase. The mineral type of AFt mainly including iodine remained in the altered cement solid along L/S and finally released iodine by dissolution in large L/S. It was confirmed that iodine was completely released at 1400 in cumulative L/S. Based upon these findings, the iodine release from this cement solid was evaluated by the solubility equilibrium model. The alteration of minerals in the cement and the release of iodine during immersion were evaluated in thermo-equilibrium conditions by using the geochemical calculations code PHREEQC. The calculated concentration of iodine and mineral phase were compared to the results of the immersion tests. Iodine release behavior consistent with mineral phases could be interpreted under a hypothesis, in which precipitation rate of iodine into the most thermo-dynamic stable phase was so low that the other reactions could occur first. More realistic conditions, such as fresh groundwater with some chemical components, were also studied and found that iodine will be confined long enough for the requirement. This research is a part of “Research and development of processing and disposal technique for TRU waste containing I-129 and C-14 (FY2011)” under a grant from the Japanese Ministry of Economy, Trade and Industry (METI).

Halide-rich waste streams, such as those arising from the pyrochemical reprocessing of plutonium, pose particular problems for immobilisation. The solubilities of these anions in silicate melts are generally very low and their inclusion (particularly of Cl) can have substantial detrimental effects on the properties of the glass formed. Therefore conventional vitrification of these wastes is not suitable for their immobilisation and disposition. As alternatives to vitrification, calcium chlorosilicate and quadridavyne, two natural mineral phases containing substantial concentrations of Cl, are being investigated as potential ceramic matrices for the immobilisation of these wastes. Solid solutions doped with surrogate waste have been fabricated via conventional solid state methods at relatively low temperatures in order to minimise the loss of Cl through volatilisation. In the case of calcium chlorosilicate, characterisation by XRD, SEM and DTA shows a single phase product with high Cl retention, which can be either pressed and sintered or encapsulated within a glass matrix to form a monolithic waste-form.

Apatites are an abundant group of minerals, important in a wide variety of applications. This is partly facilitated by their considerable compositional flexibility. For example, they can accommodate a very wide range of species, including those that cannot be incorporated in currently employed nuclear waste hosts. Molecular dynamics simulations, used in conjunction with a set of classical pair potentials, have been employed to investigate the transport of fluorine in fluorapatite. A new coupled interstitial migration mechanism is identified with a migration activation energy of 0.55 eV in the temperature range 1100-1500 K. A full description of the mechanism is provided, which differs markedly from previously proposed vacancy mechanisms for fluorine transport. Furthermore, a discussion chlorine and hydroxy ion migration in chlorapatite and hydroxyapatite respectively, is also discussed.

Tc-99 is one of the most important isotopes likely to be disposed of in the proposed UK Geological Disposal Facility (GDF) for higher-activity radioactive wastes, due to its long half-life, high fission yield and ability to migrate through the geosphere as the pertechnetate ion. However, much of the technetium is likely to be in the lower oxidation state of Tc(IV) due to the low Eh in the near field. Batch sorption experiments across the pH range have been performed on Tc(IV) using Tc-95m as a spike in the presence of silicate, iron and clay minerals. Tc(IV) solutions were used at trace concentrations to avoid precipitation as TcO2. Values for the partition coefficient (Rd) were found to range from 7 to 2e5 ml/g. Rd was heavily dependent on pH in all cases, with the highest values being found in the circumneutral area. These data will inform the performance assessment for the behaviour of technetium in the near-field of the UK&’s planned higher-activity wastes GDF. Surface complexation modelling of the data has been performed.

The preferred UK option for managing higher activity radioactive wastes is storage in a deep Geological Disposal Facility (GDF). This may then be backfilled with a cementitious material and highly alkaline porewater will develop. Cement mineral phases will act as buffers and maintain the pH at 12.5 or above for ca. 10000 years. Tc-99 is one of the most important isotopes to be disposed of due to its long half-life/high fission yield, and ability to migrate in its oxidised form. In the past, Tc was discharged to sea. It was originally thought to disperse widely, but was discovered to concentrate in seaweed. Hence, treatment with TPPB is used to precipitate it out to prevent marine discharges. This leads to the possibility that a Tc-containing floc may be sent to a GDF. However, TPPB degrades by alkaline hydrolysis at high pH, and is also prone to radiolytic degradation. Organics will be present as waste components, e.g. isosaccharinic (ISA) and gluconic acids formed by the degradation of cellulose. These are highly complexing and can cause significant increases in radionuclide solubility. A GDF will not be heterogeneous with areas of reducing and oxidising potential. This heterogeneity could mean that both Tc(VII) and Tc(IV) are present in a GDF. If Tc(VII) migrates into a low Eh area, the organics may complex with Tc during reduction to form water-soluble complexes. Also of relevance is the possibility of increased solubility when organics are in contact with Tc(IV) oxide; i.e., does the presence of organics affect the reduction of Tc(VII) to Tc(IV)? Therefore, studies have been undertaken in which Tc(VII) was reduced with and without ISA, gluconic acid, EDTA, NTA and picolinic acid, to determine whether they caused an increase in Tc solubility when TcO2 was contacted with them. In the presence of ISA and gluconic acid a lowering of [Tc(aq)] took place on reduction, showing such ligands did not prevent reduction occurring. If the reduction was to Tc(IV), then the final aqueous concentration should be the same as that produced by the addition of the same ligands to Tc(IV) solution. However, the final Tc solubility in the system where reduction took place in the presence of gluconate was higher than when TcO2 was the starting point. This indicates that Tc(VII) may not have been reduced to Tc(IV) but an intermediate oxidation state complex may have formed, an idea known from Tc-99m radiopharmaceuticals. The anthropogenic ligands EDTA and picolinic acid are used as decontamination agents and will find their way into intermediate level waste (ILW). They could then complex with Tc(IV), raising its aqueous concentration, and hence increasing its mobility in the cement porewaters and beyond. The conditional stability constants (measured in 0.3 M NaOH) have been determined to be log β = 26.2 ± 0.6 for Tc-EDTA, and log β = 26.9 ± 0.1 for Tc-PA. However, the overall effect of these ligands on the solubility of Tc is quite low in such systems.

Technetium-99 is long-lived (half-life is over 210,000 years) artificial radionuclide accumulated in spent nuclear fuel. It is very mobile under oxidizing conditions in geological environment. Development of durable Tc waste form is considered in many countries. Solid nonselective sorbent Layered Hydrazinium Titanate, LHT-9 (PCT/EP2010/001864) with general formula (N₂H₅)_1/2Ti_1.87O₄xH₂O was proposed for Tc sorption from aqueous solution followed with precipitate&’s conversion into durable titanate ceramic. Experiments on static sorption of technetium on LHT-9 were carried out in neutral aqueous solution of 2 g/l KTcO4 using varied quantity of LHT-9 (1-200 g/l). The highest distribution coefficients Kd of technetium (during 24 hours) were observed for 200 g/l LHT -9 (Kd=179712 ml/g), 20 g/l LHT (Kd=96737 ml/g), and 10 g/l (Kd=3929 ml/g). Another experiments on static sorption of radionuclide on LHT-9 were carried out using solutions of KTcO₄, NH₄TcO₄, NaTcO₄, RbTcO₄, CsTcO₄, and SrTcO₄ at varied pH=4, 7, and 9. Technetium containing in solutions was 1/3 of weight of LHT-9. It was found that uptake of technetium doesn&’t depend on pH and chemical composition of solution. After sorption the precipitates obtained were calcined at 400° and 800° C and studied using X-ray powder diffraction. Tc-containing titanates phases were observed. Phase and chemical composition of synthesized powders are discussed.

Technetium is a fission product produced during the burning of nuclear fuel and is particularly hazardous due to its long half life (210000 years), relatively high content in nuclear fuel (approx. 1 kg per ton of SNF), low sorption, and high mobility in aerobic environments. During spent nuclear fuel (SNF) reprocessing Tc is released either as a separate fraction or in complexes with actinides and zirconium. Although Tc has historically been discharged into the marine environment more stringent regulations mean that the preferred long term option is to immobilise Tc in a highly stable and durable matrix. This study investigated the feasibility of incorporating of Tc analogues (Re, Mo) in various crystalline host matrices, prepared by solid state synthesis, under different atmospheres. Samples have been characterised with X-ray and electron diffraction, and scanning electron microscopy. As expected the solid solubility of Re and Mo was shown to be dependent on processing atmosphere.

The DOE Fuel Cycle Research & Development (FCR&D) Program is developing aqueous and electrochemical approaches to the processing of used nuclear fuel that will generate technetium (Tc) -bearing waste streams. Metallic alloy waste forms are being developed for these waste streams. Laves-type intermetallics have been observed to be the dominate phase in the alloy compositions being designed for the immobilization technetium. These phases include hexagonal C14 with the composition (Fe,Cr)2Mo, cubic C15 phase for the (Fe,Ni)2Zr composition, and the Pd2Zr phase which was identified as a hexagonal close-packed structure. The occurrence of these phases in the proposed alloy nuclear waste form demonstrates their importance for understanding and modeling the long-term potential release behavior of technetium under disposal conditions. The hexagonal C14 Laves phase is considered to form first and contains Tc. It appears that the cubic C15 phase then forms. PdZr2, also a close packed Laves phase, segregates out within the bcc-iron phase that solidifies interstitially last of all. The C14 and C15 Laves phases are close-packed intermetallic structures. The C14 structure is hexagonal with the MgZn2-type structure. Because of the similarity in the structures, layering may occur as the phases can shift from one polytype to another with only minor changes in composition. For instance, the hexagonal layers in the C15 structure are along [111], while similar stacking is along [0001] in C14. This type of complex layering was observed in the Tc-bearing intermetallic phases.

Technetium-99, a β-emitting radioactive fission product of 235U, formed in nuclear reactors, presents a major challenge to nuclear waste disposal strategies. Its long half-life (2.1 x 10^5 years) and high solubility under oxic conditions as the pertechnetate anion [Tc(VII)O4-] is particularly problematic for long-term disposal of radioactive waste in geological repositories. In this study, we demonstrate a novel technique for quantifying the transport and immobilisation of technetium-99m, a γ-emitting metastable isomer of technetium-99 commonly used in medical imaging. A standard medical gamma camera was used for non-invasive quantitative imaging of technetium-99 during co-advection through quartz sand and various cementitious materials commonly used in nuclear waste disposal strategies. These include: crushed ordinary portland cement (OPC); OPC combined with blast furnace slag (BFS) or pulversised fly ash (PFA); and Nirex Reference Vault Backfill material. Pulse-input experiments of approximately 15MBq 99mTc were conducted under saturated conditions and at a constant flow of 0.33ml/min. Dynamic gamma imaging was conducted every 30s for 2 hours. Spatial moments analysis of the resulting 99mTc plume provided information about the relative changes in mass distribution of the radionuclide in the various test materials. 99mTc advected through quartz sand demonstrated typical conservative behaviour, while transport through the cementitious materials produced a significant reduction in colloid centre of mass transport velocity over time. BFS-containing cement was shown to be most effective at immobilising 99mTc, with up to 50% of the injected activity retained irreversibly by the cement. Concurrent batch experiments using 99Tc and rhenium, in conjunction with PHREEQC reactive transport modelling suggest that technetium is immobilised by Fe and S within the BFS cement. Gamma camera imaging has proven an effective tool for helping to understand the factors which control the migration of radionuclides for surface, near-surface and deep geological disposal of nuclear waste.

The stability of Tc and Ru in TiO2-rutle as both simple substitutional defects and defect clusters in solid solution has been investigated. Density functional theory based calculations have been used to confirm the solubility of Tc into the rutile structure, but only as a Tc=Tc dimer. Single Tc defects were found to have a positive solution energy. The transmutation of Tc to Ru is then discussed and calculations have un-covered an interesting consequence of the transmutation process. Both Ru as a single defect and in a defect cluster with either Ru or Tc has a positive solution energy indicating that the transmutation of Ru will lead to secondary phase formation. The work is then concluded with a number of calculations that suggest possible dopants that could be added to the rutile phase to prevent secondary phase formation after the Tc transmutation.

IONSIV IE-911 has been widely applied in the nuclear industry as an inorganic ion-exchanger to separate ¹³⁷Cs from waste streams due to its excellent selectivity and high thermal/radiation stability. It is a commercial mixture of a crystalline silicotitanate (CST) with the formula of (H₃O)_xNa_y(Nb_0.3Ti_0.7)₄Si₂O₁₄#9679;zH₂O - where x~2, y~1 and z~4, and a Zr(OH)₄ binder. To manage the spent ion exchanger, hot isostatic pressing (HIPing) is being applied as a route for densifying and consolidating the materials to produce a monolithic wasteform prior to final disposal. In this study, IONSIV was firstly ion-exchanged in aqueous CsNO₃ to yield Cs-IONSIV and then HIPed at 1100 °C for 2 hrs (190 MPa, Ar gas) within a mild steel can. During the HIP process, Cs-IONSIV was thermally decomposed and converted to two major Cs-containing phases, Cs₂TiNb₆O₁₈ and Cs₂ZrSi₆O₁₅, and a series of other phases. The microstructure and phase assemblage of the HIPed samples as a function of Cs content were examined using XRD, XRF, SEM, and TEM-EDX. The leaching and durability of HIPed IONSIV was also investigated using the MCC-1 and PCT-B standard test methods. These show very low Cs leach rates and the promise of safe long-term immobilisation of Cs from IONSIV as well as suggesting these phases are superior to hollandite - the material targeted for Cs sequestration in SYNROC.

Hollandites, which have the general formula (Ba_xCs_y)(Ti,Al,Fe,Mg)₈O₁₆ (x+y < 2), possess a three-dimensional framework structure of (Ti,Al,Fe,Mg)O₆ octahedra, via edge- and corner-sharing, with Ba and Cs occupying the tunnel sites. The flexibility of this framework for accommodating both Cs⁺ and Ba²⁺ is particularly useful for ¹³⁷Cs immobilization, as ¹³⁷Cs transforms to ¹³⁷Ba through beta decay with a half-life of about 30 years and this framework flexibility ensures the stability of hollandites over the decay period. In this study, we synthesized a series of hollandite phases via combustion of metal citrates. In situ neutron and synchrotron X-ray diffraction experiments were conducted to interrogate their crystal structures at high-temperature and/or high-pressure conditions. Rietveld analysis of the obtained data allowed determination of lattice parameters, atomic positions and atomic displacement parameters as a function of temperature and pressure. The bulk moduli, thermal expansion coefficients and other thermo-mechanical properties have thus been obtained. Lastly, the enthalpies of formation of hollandites from their constituent oxides and elements were measured using high-temperature oxide-melt calorimetry. The determined thermodynamic stability is discussed in terms of crystal chemistry.

Nuclear waste disposal is a significant technological issue, and the solution of this problem (or lack thereof) will ultimately determine whether nuclear energy is deemed environmentally friendly, despite significantly lower carbon emissions than fossil fuel energy sources. A critical component of any waste disposal strategy is the selection of the waste form that is tasked with preventing radionuclides from entering the environment. The design of robust nuclear waste forms requires consideration of several criteria, including: radiation tolerance, geological interaction and chemical durability; all of these criteria ensure that the radionuclides do not escape from the waste form. However, relatively little attention has been paid to the phase stability, and subsequent durability, of candidate waste forms during the course of daughter product formation; that is, the chemical aging of the material. Systematic understanding of phase evolution as a function of chemistry is important for predictions of waste form performance as well as informing waste form design. In this presentation, we highlight the research challenges associated with understanding waste form stability when attempting to systematically study the effects of dynamic composition variation due to in situ radionuclide daughter production formation. These challenges will be presented in the context of recent experiments and atomic scale simulations performance on isotopically pure samples.

During the past forty years, the materials science of nuclear waste forms has focused on the stability and long-term behaviour on nuclear waste glasses and used nuclear fuels, mainly UO2. During this same period, substantial quantities of Pu, now more than 2,000 metric tones, have accumulated, either still in the used nuclear fuel or chemically separated for weapons or energy applications. This "excess" plutonium, as well as associated "minor" actinides (Np, Am and Cm) offer a new, but seldom pursued, opportunity for the safe geologic disposal of transuranium elements. A variety of materials, with mineral analogues, including oxides, silicates and phosphates, have been investigated because of their high capacity to incorporate actinides, their chemical durability, and in some cases, their resistance to the radiation-induced transformation to the aperiodic state. There has been substantial interest in isometric pyrochlore, A2B2O7 (A= rare earths, actinides; B = Ti, Zr, Sn, Hf), for the immobilization of actinides, particularly plutonium. Systematic studies of rare-earth pyrochlores have led to the discovery that certain compositions (B = Zr, Hf) are stable to very high doses of alpha-decay event damage. Three different processes have been observed: i) radiation-induced amorphization, ii) an order-disorder transformation and iii) phase decomposition. The radiation stability of these derivatives of the fluorite structure-type is closely related to the structural distortions caused by compositional variations that affect electronic structure and bond-type. Based on this very fundamental understanding of the radiation response, durable, actinide waste forms can be designed for specific temperature and radiation dose conditions, such as those found in very deep boreholes.

A calcium phosphate ceramic waste-form has been developed at AWE for the immobilisation of chloride containing wastes arising from the pyrochemical reprocessing of plutonium. In order to determine the long term durability of the waste-form, aging trials have been carried out at PNNL. Ceramics were prepared using Pu-239 and -238 and aged for up to 5 years. Samples were characterised by PXRD at regular intervals, modified Materials Characterisation Centre (MCC-1) tests after approximately 1.5 and 5 yrs, and Single Pass Flow Through (SPFT) tests after approximately 5 yrs. While PXRD indicated no in-growth of new phases over this time, although some damage was detected in the Pu-238 samples after exposure to 2.8 x1018 α decays. Release rates of constituents from MCC-1 tests of the Pu-239 samples after 1.5 and 5 yrs were very similar. Dissolution rates of the Pu-238 samples were also similar before aging and after aging, although these were somewhat higher than values obtained for the Pu-239 ceramics. However, in the SPFT tests no significant difference in the release rates observed between the Pu-238 and -239 samples, indicating that the radiation induced damage did not have a significant effect on the dissolution rates. Therefore, it is suggested that the observed differences in the MCC-1 tests for the Pu-238 and -239 samples arose from radiolysis effects.

Zirconate ceramics with the pyrochlore and related fluorite structures are crystalline matrices with potential to host plutonium and minor actinides as durable waste forms for long-term geologic disposal or as inert matrices for actinide transmutation. Members of the zirconate pyrochlore system are not only chemically durable with release rates typically < 10^-5 gm^-2d^-1 based on zirconium, but have also been shown to have remarkable resistance to amorphization under ion-beam irradiation. In this work we investigate the incorporation of plutonium in pyrochlore-structured La₂Zr₂O₇, with different compositions e.g., La_1.9Pu_0.1Zr₂O₇ (targeting Pu³⁺) and La_1.8Pu_0.1Ca_0.1Zr₂O₇ (targeting Pu⁴⁺), in some cases using alkaline earth metals for charge compensation. The samples were prepared by a modified alkoxide route using stoichiometric amounts of tetrabutyl zirconate and an aqueous solution containing the nitrates of Ca, Sr, Pu, and La. After powder preparation, a pellet for each sample was pressed and sintered in air, argon or H₂/N₂ at 1500°C for 24 hours. The samples were then characterized by X-ray powder diffraction (XRD) and scanning electron microscopy (SEM) and the plutonium oxidation state was investigated by diffuse reflectance spectroscopy (DRS) and X-ray absorption near edge structure (XANES) spectroscopy. All samples were identified by laboratory XRD as single phase materials with the pyrochlore structure, with the exception of La_1.9Pu_0.1Zr₂O₇ (prepared in H₂/N₂), which showed ZrO₂ as a minor impurity phase. This was confirmed by backscattered SEM analysis. The SEM micrographs also showed that the sample with composition La_1.9Pu_0.1Zr₂O₇ (sintered in H₂/N₂) to be quite porous while the remaining air sintered samples were denser. The Plutonium-oxidation state in each sample was determined using DRS. For all the air and argon sintered samples, the DRS showed peaks characteristic of Pu⁴⁺. These peaks were not present in samples sintered in H₂/N₂, and the DR spectra were rather featureless. The plutonium oxidation state was further confirmed using XANES analyses.

The formation of radioactive waste from nuclear power reactors will be briefly discussed. Synroc was originally a titanate ceramic designed for immobilisation of reprocessed nuclear fuel, but more recently the ANSTO synroc group has focussed on the use of hot isostatic pressing to consolidate glass, glass-ceramics or ceramics for immobilisation of a range of high- and intermediate level nuclear wastes. The prime mission of the group is currently the design of a plant to immobilise intermediate level liquid waste arising from the production at ANSTO of 99Mo for radiopharmaceutical purposes. However basic work on waste form science is also undertaken at ANSTO, with studies on radiation self-damage phenomena, actinide valences in candidate waste forms, dealing with halide-bearing nuclear waste, and a range of other wastes. Geopolymers as immobilisation candidates for intermediate level will also be discussed.

Numerous processes from nuclear facilities generate important volume of radioactive effluents which should be treated in order to minimize their impact on environment. Among those, gamma emitter 137Cs is one of the most abundant fission products. Bulk cyano-bridged coordination polymers based on hexacyanometallates of transition metal called also Prussian Blue analogous are a long time known for their ability to selectively cesium ions capture over a wide range of pH even in saline solutions. This property comes from the insertion of cesium ions inside the crystalline three dimensional structures of cyanometallates, and/or also by an ionic exchange mechanism. In France, the industrial treatment of 137Cs contaminated liquid wastes was done thanks to a reliable and safe process based on Potassium Nickel Hexacyanoferrate co-precipitation process. This process is simple and cheap but does not allow an easy column treatment of large volume of effluents and generate a large quantity of sludge to be confined. Then elaboration of innovative materials able to remove radioactive Cs with a continuous process and minimize the waste volume, matching with the classical waste confinement matrix such as cement or glass, is a challenge. Thus, together with the study of these bulk selective sorbents, we developed the design of hybrid materials so that they can be used in continuous mode like cartridge process and that they can easily permit an efficient confinement. In the present work, we present first comparative data of the Cesium sorption capacity and selectivity obtained with different bulk cyano-bridged coordination polymers. Then we propose an easy and cheap way of the synthesis of hybrid materials containing cyano-bridged coordination polymer nanoparticles covalently linked to porous silica pearls. Cs sorption experiments on the obtained nanocomposites were performed in batch experiments in order to evaluate their sorption capacity and selectivity in different solution. The obtained maximum adsorption capacity (Qmax) of the composite (0.1 mmol/g) evaluated in pure water is lower than the bulk one (0.4 mmol/g). However Qmax calculated in mmol per gram of Co3[Fe(CN)6]2 nanoparticles loaded in the silica support (1.3 mmol/g) is three times higher in comparison with the bulk one. The distribution coefficient -defined as the equilibrium ratio between the quantity of the adsorbed on solid and the remaining in solution cesium - obtained in radioactive sea water is similar to the one obtained by the industrial co-precipitation process using pre-formed particles. Then experiments in column demonstrate that these silica pearls nanocomposite are more efficient for in-flow Cesium removal compared to batch process and show the high potential of these nanocomposite materials. In addition, after the extraction of radioactive cesium, these contaminated silica pearls can act as confinement matrix by closing the porosity of the silica matrix.

In the forthcoming years, the needs in uranium are expected to increase significantly. In addition, in the perspective of sustainable development, the exploitation of uranium ores, including leaching, requires to be optimized. In this purpose, reliable thermodynamic data are needed to forecast the long term behavior of uranium during leaching or decommissioning steps. Apart from uraninite/pitchblende, uranium phosphates and vanadates, including torbernite Cu(UO₂)₂(PO₄)₂.8-12H₂O and carnotite K₂(UO₂)₂(VO₄)₂.3H₂O, are present in deposits of economic interest. This work was then focused on thermodynamics of the {P₂O₅-V₂O₅-UO₂} system, which remains widely unknown. In this aim, the relevant phases were first synthesized and exhaustively characterized. For the carnotite, a dry route was favored, while torbernite was obtained through a wet chemistry method. All the solids obtained were then characterized by IR and µ-Raman spectroscopies, DTA-TGA, XRD and ESEM then compared to natural samples. The solubility data were then reached through two approaches, i.e. over-saturation (precipitation) and under-saturation (dissolution) conditions. In both cases, regular sampling of the solution was performed to monitor the elementary concentrations versus time through ICP-OES measurements. The first points associated to the evolution of elementary concentrations provided information on the reaction kinetics while concentrations determined at equilibrium were introduced into the CHESS software to determine the activities of each species, using the Davies&’ model. Such data further allowed to calculate thermodynamic data (Δ_RH°, Δ_RG°, Δ_RS° and K_S,0°). For torbernite, K_S,0° was evaluated around 10^-52 at room temperature from over-saturation experiments, in good agreement with the scarce data reported in the literature. The experiments performed at 60 and 90°C then allowed to calculate K_S,0°(T) (respectively around 10^-49 and 10^-48) and a first value of Δ_RH° = 114 ± 16 kJ/mol. Moreover, such values were supported by under-saturation studies performed in various media (1M HCl, HNO₃ or H₂SO₄). The dissolution being always congruent, the K_S,0° values were calculated for the three media. For each media, it reached around 10^-52, very close to the value obtained in over-saturation. In addition, first experiments in under-saturation conditions made on carnotite led to a first value of K_S,0° of 10^-64 for the three media. However, complementary experiments (with various temperatures and acid concentrations) are currently under progress to allow the evaluation of Δ_RH°. Since such methodology was successfully applied to phosphate and vanadate compounds, it will be developed in the near future for other phases as well as for natural samples to study the effects of both chemistry, structure and microstructure on the dissolution of such compounds.

The amount of accumulated irradiated graphite is already huge and is gradually growing. The main source of irradiated graphite radioactive waste is from uranium-graphite reactors which undergo decommissioning and which have used graphite for moderation and reflection of neutrons. There are more than 100 of such type facilities that are mainly located within United Kingdom, France, countries resulted from disintegration of former USSR, USA and Spain. There are not yet developed either technical solutions or industrial technologies to immobilise the radioactive and contaminated with nuclear fuel inclusions graphite. Flameless molten salt oxidation (MSO) of waste is one of prospective methods to treat the irradiated graphite [1,2]. MSO technology does not require fine grinding of irradiated graphite which is an advantage. Molten salts are able to retain a considerable part of radionuclides, to neutralise acidic gases, moreover spent salts can be vitrified on completion of decontamination process. We have used the thermodynamic modelling code TERRA [2] to simulate the MSO decontamination process and assess its efficiency for various salt systems. Equilibrium compositions of both condensed and gaseous phases were calculated on changing the content of irradiated graphite as a function of processing temperature. Laboratory tests were carried out aiming to decontaminate the irradiated graphite by removing the outer layers on graphite blocks using non-complete MSO. We have used for tests real irradiated and radioactively contaminated graphite sleeves of nuclear reactor IR AM. As basic molten salt baths for MSO we have used lithium, potassium and sodium carbonates. Sodium sulphate, boron oxide and barium chromate were used as oxidising media. The experiments were carried out in the temperature interval 600-1000 C. The efficiency of graphite decontamination has been controlled based on measurement of residual radioactivity of Cs-137 and Co-60 in the tested samples after MSO decontamination. Obtained data from experiments have demonstrated the feasibility of MSO decontamination of irradiated graphite based on irradiated graphite near surface layer oxidation. The oxidation rate and decontamination efficiency do mainly depend on oxidiser used and processing temperature. [1] Gay R. L., Rockwell International Corporation, Method for disposing of radioactive graphite and silicon carbide in graphite fuel elements, patent US 5.449.505, September 12, 1995. [2] A.A. Romenkov, M.A. Tuktarov, L.I. Minkin, V.P. Pyshkin, Environmental Safety, 3 (2006) 44. [3] Trusov B.G. Trudy GUP MosNPO “Radon” 13 (2007) 21.

In an integrated treatment/recycling fuel cycle, a key-step concerns the transition between i) the purified actinides in solution (mainly Pu and U) following an elaborated hydrometallurgical treatment of the used fuel and ii) the mixed actinide oxide used in the fabrication of a fresh fuel, implying the co-conversion of these recycled actinides into a ceramic precursor. Recently, many research works have been devoted to the structures of actinide co-precipitates, gels or other forms, which represent the first solid state phase of the recycled actinides after fuel dissolution and liquid/liquid extraction partitioning steps. The present contribution describes ongoing research on the thermal treatment of these precursors to produce mixed oxides. New insights into the reaction mechanisms of the conversion of actinide oxalates into oxide are principally given. The reactivity of these systems is then briefly compared to other co-conversion routes such as thermal co-denitration and calcination of mixed gels or other solid-state precursors. This comparison focuses on the potential of these co-conversion methods to produce (U,Pu)O2 solid solutions or UO2/PuO2 mixtures prior to the sintering step of the fuel material fabrication. It discusses the intermediate steps affecting the characteristics of the oxide end-product. It emphasizes the importance of the atmosphere imposed to the powder during the thermal treatment and some key-aspects of the secondary reactions intervening between the solid phase and the gaseous by-products. Moreover, the actinide chemistry, in particular redox phenomenon, is of primary importance for particular systems. This contribution highlights the relative lack of exhaustive data dealing with the calcination of actinide solids into oxides. This transformation has often been considered as a “black box” ending the treatment even though it often consists in an important preliminary step of the fuel fabrication within a closed fuel cycle. Next nuclear generation systems, where multi-recycling of fissile and fertile materials is put forward, motivate the acquisition of extended basic data in this field to better integrate used fuel treatment and fresh fuel fabrication.

Advanced burner reactors are designed to reduce the amount of long-lived radioactive isotopes in spent nuclear fuel. The input feedstock for advanced fuel forms derives from either recycled light water reactor fuel, or recycled fast burner reactor fuel. In order to achieve higher performance and increase operational safety, these advanced reactors require novel fuel concepts, made from new materials. One promising pathway to improve fuel performance is the creation of metal or ceramic scaffolds, into which fuel may be placed with greater precision than in existing CERMET fuels. In this presentation, the design and manufacture of novel structures by “freeze casting” will be described. Freeze casting (or “ice templating”) is a directional solidification process ideal for the production of both metal and ceramic fuel scaffolds. This process inherently allows for the manufacture of a range of custom-tailored fuel pellet designs. The mechanical, thermal, and neutronic properties of both metal and ceramic scaffolds will be compared and contrasted, in order to shed light on the performance and lifetime behavior of these novel fuel designs.

Sodium zirconium phosphate (NZP) type ceramics accommodate approximately 42 elements of the periodic table including most fission products derived from nuclear power plant fuel. As such, NZP-structure type ceramics have considerable potential as host materials for the immobilization of radioactive waste as well as candidate inert matrices for minor actinide burning. It is therefore important to investigate the behaviour of this material under irradiation conditions in order to verify its long-term stability. In this study strontium zirconium phosphate (an NZP-type structure ceramic) has been irradiated with gold and helium ions to simulate the consequences of alpha decay. The effects of the irradiation on the structural as well as macroscopic properties (e.g. density and hardness) are investigated using grazing-incidence X-ray diffractometry, Raman spectroscopy, scanning electron and atomic force microscopy, and nano-indentation. Irradiation by gold ions results in significant changes to the crystalline structure and hardness. After a fluence of 10¹⁵ gold ions/cm², strontium zirconium phosphate undergoes structural amorphization, a volume reduction, and an increase in hardness. These results as well as the results from He-ion irradiation are discussed with regard to the application of NZP-structure type ceramics as inert matrices for minor actinide burning or as host materials for the immobilization of radioactive waste.

Fergusonite and scheelite-structured ABO4 ternary oxides are an important class of materials owing to their technological applicability and geological significance. In spite of their growing interest as potential wasteform ceramics, only very little is known about their behaviour under irradiation in regards to other ABO4 analogues such as zircon and monazite. To this purpose, we have studied and compared the effects of ion-beam irradiation on compounds LaVO4, YNbO4 and CaWO4 by 1 MeV Kr+ ions as a function of irradiation temperature (50 - 600K). Resulting critical temperatures for amorphisation (Tc) differ slightly for LaVO4 and YNbO4 each with a Tc of 400K and 450K respectively. CaWO4 shows stronger amorphisation ‘resistance&’ and has a Tc of 200K. The susceptablity toward amorphisation and disorder in each structure is discussed in terms of their structural parameters as well as the stopping powers, displacement energies, and defect energies of the materials. The phase transitions that occur between tetragonal scheelite and monoclinic fergusonite will also be highlighted.

This study presents structural analysis of crystalline and radiation-damaged zirconolite, CaZrTi₂O₇, and pyrochlore, Gd₂Ti₂O₇, both potential actinide-accommodating nuclear waste materials, using molecular dynamics (MD) and connectivity topology analysis - a powerful method for describing both crystalline structures and their metamict or amorphous analogues, because it places no reliance on symmetry operators or periodic translation, both of which v