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 vanish upon introduction of disorder to a material. The work establishes characteristic topological differences in the connectivity of each structure and finds evidence that amorphization induced by alpha-recoil displacement cascades still retains certain short- and intermediate-range ordered configurations, particularly for Ti atoms. [TiO_x] polyhedral edge-sharing chains are observed in the metamict state in both materials, which may act to stabilize the radiation-damaged structure and prevent recovery of the initial crystalline phase. We also present an assessment of the predicted amorphizability of zirconolite based on the topological constraints imposed by its structure, finding that the varying structural rigidity of the layers in the structure is crucial to its amorphizability potential. The hexagonal tungsten bronze structure [TiO_x] layer in particular provides weak constraints that are responsible for zirconolite&’s comparative ease of amorphization.

Molecular dynamics simulations using thermal spikes have been carried out to investigate the effect of pressure on the time-dependent generation of defects under irradiation in the three common polymorphs of TiO₂: rutile, anatase and brookite. The effect of crystal structure on the tollerance to radiation damage was first investigated, highlighting the experimental observation that the rutile phase is the most tollerant to damage. The density of the phases was then varied and the same thermal spike methodology repeated with some interesting results suggesting a strong correlation between density and radiation tollerance.

In the realm of radioactive nuclear materials, a major challenge to the development of materials is the measurement of the properties for which the material is being developed. For example, the phenomenon of microstructure evolution of a nuclear fuel in reactor is well known but the details of the effects of the change on the behavior of such important issues as thermal conductivity, mechanical properties, and phase formation have not been quantified at the grain size level. There is a strong need to develop or adapt advanced instrumentation for measurements on radioactive materials. Idaho National Laboratory has an ongoing effort to develop or adapt a variety of measurement techniques to highly radioactive materials. A laser based device termed the Scanning Thermal Diffusivity Microscope, conceived and developed over the past few years, has recently been installed in a hot cell where examinations of fresh and irradiated fuel samples have begun in order to profile the thermal diffusivity of fuels and materials at 50µm spatial resolution. A second generation instrument close to implementation will soon give thermal conductivity values at 5-10 µm. The unique application of dual-beam focused ion beam (FIB) to the preparation of highly radioactive material samples has become an exceedingly useful tool for determining 3D grain orientation (EBSD), mechanical properties by nano/micro indentation or compression testing, microstructure through transmission electron microscopy, and nano-scale element distribution by atom probe tomography. This contribution will present the current state of the implementation plan of these instruments to highly radioactive fuels and materials and examples from ongoing irradiated fuels and materials studies will be given.

Under ion irradiation, all crystalline materials display some degree of dynamic annealing when defects experience evolution after the thermalization of collision cascades. The exact time and length scales of such defect relaxation processes are, however, unknown even for Si at room temperature. Here, we propose a method to measure effective diffusion lengths and relaxation times of mobile defects that dominate the formation of stable post-irradiation disorder. A defect lifetime of about 5 ms and a characteristic defect diffusion length of about 30 nm are measured for Si at room temperature, essentially independent of the average density of ballistic collision cascades. Defect relaxation appears to be dominated by a second order kinetic process. We discuss implications of these findings for the development of predictive models of radiation damage buildup in solids. This work was performed under the auspices of the U.S. DOE by LLNL under Contract DE-AC52-07NA27344.

A series of static lattice calculations were performed to determine the site selectivity of cations of differing size and valence when substituted onto the Ca sites of the calcium chlorosilicate (Ca₃(SiO₄)Cl₂) lattice, a potential host phase for the immobilisation of halide-rich wastes arising from pyrochemical reprocessing operations. Atomic scale simulations indicate that divalent cations are preferentially substituted onto the Ca1 site, whilst tri- and tetravalent cations are preferentially hosted on the Ca2 site, with the Ca1 site favoured for forming the vacancies necessary to charge-balance the lattice as a whole. Multi-defect calculations reveal that the site selectivity of the dopant cations is dependant on their ionic radii, as the ionic radii of the divalent cations increases, substitution onto the Ca1 site becomes more and more strongly favoured whereas the inverse is true of the trivalent cations, where substitution onto the preferred Ca2 site becomes more strongly favoured as their ionic radii decreases.

This research investigates the temperature processing of the NPP and radiochemical facilities spent ion-exchange resins (IER). This method allows processing initial IERs into the solid bulk material suitable for its further incorporation into stable solid matrices (e.g. cement or glass). The preliminary thermographic analysis (TGA) has been performed to determine boundary values of the thermal IER decomposition process. TGA has revealed that the maximum IER processing temperature should not exceed 300-500°C depending on a kind of resin since a further temperature increase will result in the IER decomposition causing the formation of toxic combustion products. The experiments on determination of conditions for temperature processing of ion-exchange resins were conducted using the pilot installation based on the rotary calciner design. The basic working unit of the rotary calciner is a stainless steel tube-retort tilted slightly with respect to the horizon. A motor-reducer is used for the retort rotation. The retort is heated by SiC heaters. The processes of water evaporation, IER suspension drying and bulk powdery product formation are successively taking place as the initial material is passed through the calciner retort. The experiments on temperature IER processing were conducted using KU-2 cation resin at a solid-to-liquid ratio of 1:5. The rate of the suspension feed into the calciner and the rate of the retort rotation were varied in the ranges of 5-10 l/h and 10-20 r/min, respectively. The processed IERs (PIERs) were analyzed to determine the following physicochemical characteristics: particle-size distribution, bulk density, and angle of repose. It has been shown that as a result of temperature processing the volume of the initial IER with a natural moisture content decreased by 3 times, while the final moisture content of IER (~ 3 wt.%) remained virtually invariable when kept in the open air for 7 days. So good friability and low residual moisture allow to store IER processing products in the metal drums, hereinafter placing it in nonreturnable concrete containers. It is also possible to fill the IER processing product into concrete containers directly without an additional packaging (drums).

The backfill is a very important element of the multibarrier protection of a near surface repository for low and intermediate level radioactive wastes (LILW) and is designed to perform several functions, such us a retain radionuclide migration from the repository to the biosphere, changing the characteristics of penetrating water, and providing the ability to perform the necessary actions in connection with changing in status of the repository after a specified period of storage. This paper presents results of evaluation of sorption properties of materials considered to be components of the backfill for near surface repositories for LILW. The quartz sand of local origin was offered as a major component of the backfill. Clinoptilolite, hematite, and magnesium oxide were used as additives, which can increase sorption capability of backfill to radionuclides 137Cs, 90Sr, 60Co, U and Pu. To select materials for a backfill mixture is necessary to study its sorption properties. This work focused on determination of such a parameter as degree of sorption of radionuclides from aqueous solutions by the backfill mixtures in presence of competing ions and complexing agents. The experiments were conducted under static conditions, the ratio of solid and liquid phases was 1:2 and 1:4. Solid phase was separated in centrifuge. Sorption of 137Cs in sand was determined to be a 76-99 % in the presence of stable cesium and sodium, respectively. Sorption of 90Sr varies from 62 to 86% and depends on the presence stable calcium and strontium in solution. In the presence of clinoptilolite (not more whan 20 wt%) in backfill mixture sorption of 90Sr increased almost to the sorption value of the same radionuclide in pure clinoptilolite ~ 100% and did not depend on the presence of competing ions in solution. Sorption of 60Co from aqueous solution in sand was determined to be 99 %. Presence of complexing agent in the form of ethylenediaminetetraacetic acid (EDTA) in concentration of 0,005 M/l decreased sorption of 60Co down to 10 %. Sorption of U from aqueous solution in sand was determined to be 74 %. In the presence of NaNO3 (0,01 M/l) sorption of U was determined to be 76 %. Presence of Na2CO3 (0,005 M/l) increased sorption of U up to 91 %. . In the presence of hematite in backfill mixture ( 20, 30, 50 wt%) sorption of 90Sr from solution enhanced even in the presence of complexing agent. Sorption of Pu from aqueous solution in sand was determined to be 80-84 %. In the presence of clinoptilolite and magnesium oxide in backfill mixture, sorption of Pu was determined to be 60-63 %, while sorption of other radionuclides (137Cs, 90Sr, 60Co, and U) did not reduce, and reached up to 99% in the presence of magnesium oxide in backfill mixture in amount of 5-10 wt% . Thus, these studies allow to select a good radionuclide retaining backfill for near surface repository of LILW.

The main hydrate of cement is calcium silicate hydrate (CSH). Such a cement-based material is essential for constructing the geological disposal system of TRU radioactive wastes including I-129 in Japan. So far, the sorption behavior of iodine on CSH gel has been examined by using the CSH sample dried once. However, the Japan&’s repository would be constructed under water table. Therefore, we must focus on also the interaction of altered cementitious material and iodine under the condition saturated with saline groundwater. In this study, the sorption behavior of iodine into CSH gel, formed without dried processes, was examined in imitated saline groundwater. Ca/Si ratio was set to 0.4, 0.8, 1.2 and 1.6, and NaCl concentration also was set to 0.6, 0.06 and 0.006. Each sample was synthesized with CaO, SiO2 (fumed silica), and distilled water in a given combination of 20 in liquid/solid ratio. A NaI solution was added after curing the CSH gel (hereinafter referred to as the “surface sorption sample”) for 7 days, setting the initial concentration of NaI to 0.5 mM in sample tube. The values of Eh and pH of each sample showed iodide ions as the chemical species of iodine in the sample tube. Furthermore, this study prepared the “co-precipitation sample” of CSH gel with iodine. Here, the NaI solution was added before curing the CSH gel. For all samples, the contact time-period of the CSH gel with iodide ions was set to 7 days. After each contact time-period, each sample for analyses was separated into the solid and the liquid phases by 0.20 µm membrane filter. In the liquid phase, the concentrations of Ca, I, Si and Na ions in the liquid phase were measured by ICP-AES. Besides, the Raman spectra were obtained from the solid phases of the surface sorption sample and the co-precipitation sample without dried process. The results showed that the sorption of iodine into CSH gel strongly depends on the amount of water included in the CSH gel. Such a sorption behavior was confirmed in both co-precipitation samples and the surface sorption samples, even if the Ca/Si ratio is low. This means that iodide ions can be easily immobilized through the water-molecular of CSH gel. Besides, Na concentration did not so much affect the sorption behavior of iodine into CSH gel. In addition, the Raman spectra showed that the degree of polymerization of SiO4 tetrahedrons in CSH gel was unaffected with increasing Na ions concentration. These results suggest that the CSH gel saturated with groundwater would retard the migration of iodide ions, even if the groundwater includes salinity.

Oxidizing species are expected to prevail in the vicinity of Spent Nuclear Fuel surface due to water radiolysis by the alpha radiation emitted by the wastes. Among them, oxygen and hydrogen peroxide are expected to be the main oxidant species. Besides, the anoxic corrosion of the steel containers will produce hydrogen and, according to calculations, it might accumulate in the repository to up to approximately 50 bars. The high hydrogen pressure together with the presence of potential catalyst particles in the fuel are believed to generate reducing conditions. Hence, it is of great interest to determine which of the two antagonistic effects, reducing or oxidizing, will prevail under conditions close to the ones expected in the repository. For this reason, an in situ X-ray photoelectron spectroscopy (XPS) study of surface evolution of an UO2 sample has been conducted. The XPS is connected with a reactor chamber were the solid sample can be put in contact with either Ar(g), H2(g), O2(g) or Ar saturated with hydrogen peroxide, at various pressures and temperatures. Mixtures of the different gas streams can be used as well. The solid sample was a disk fabricated by pressing freshly powdered UO2. Its final diameter and thickness were 10 and 1 mm, respectively. A total of twelve different experiments were performed. Six of them were conducted under reducing conditions, using H2 at 500 C. The other six were made under oxidizing conditions. One of them was carried out with an O2 stream while the rest were performed by using a gas stream with different proportions of hydrogen and hydrogen peroxide at 350 C. XP spectra were recorded after each reaction without exposing the sample to the atmosphere. All the spectra were analysed by deconvolution of the characteristic uranium peaks: U 4f7/2 in two different peaks, corresponding to either U (IV) or U (VI). The results obtained in this work show that under the presence of both high pressure hydrogen and representative hydrogen peroxide concentrations, the UO2 surface is oxidized.

Some of transuranium (TRU) waste packages contain considerable amount of nitrates deriving from the reprocessing process. Nitrate ion (NO₃^-) and subsequent ammonia (NH₃) and ammonium ion (NH₄⁺) may have a significant effect on the migration behavior of radionuclides by forming complexes and compounds with radioactive metal ions. Nickel (Ni-63) as radioactivated product and palladium (Pd-107) as fission product potentially form stable ammine complexes. In the present study, therefore, we focus on the sorption behavior of Ni and Pd on the pumice tuff in the presence of NH₃(aq)/NH₄⁺. Under different NH₃(aq)/NH₄⁺ concentration, pH and ionic strength, distribution coefficient (D) of Ni and Pd is determined and the obtained D values are discussed using the surface complexation model. Sample solutions containing 10^-6 mol/dm³ (M) of Ni²⁺ at pH 8 and 11 were prepared. The ionic strength was adjusted to 0.01, 0.1 and 1 by NaClO₄ and initial NH₄⁺ concentration was adjusted to 0 to 10^-2 M by NH₄ClO₄. The powder of pumice tuff (63 - 125 mu;m, 0.3 g) was added to each 30 mL sample solution as solid phase. Samples for palladium were prepared in the same method. All sample preparation was performed in an Ar glove box (O₂ gas < 10 ppm). After several weeks, 1 mL of supernatant of each sample was filtered through 0.45 mu;m syringe filter and 3k and 10k Da ultrafiltration membranes. After filtration, the Ni and Pd concentration was determined by ICP-MS, and the distribution coefficient (D) was obtained. For Ni system, the D values increased with an increase of pH. In alkaline pH region, the solubility of Ni hydroxide is low enough to precipitate, and the D values may be overestimated due to the precipitation of Ni hydroxide. Under neutral pH condition, no significant dependence of D values on initial NH₄⁺ concentration ([NH₄⁺]_ini < 10^-2 M) was observed. The prediction from thermodynamic data suggested that the dominant soluble species was Ni²⁺, not Ni ammine complex. The log D values at neutral pH were analyzed assuming that Ni²⁺ forms a surface complex. For Pd system, on the other hand, no precipitate was observed in the previous solubility experiment and it was supposed that Pd²⁺ forms a stable ammine complex (Pd(NH₃)_m²⁺) in the presence of NH₃(aq) (initial [NH₄⁺] > 10^-4 M) [1]. The obtained D values decreased with an increase of pH, and the trend was analyzed using the surface complex model assuming that Pd(NH₃)_m²⁺ sorbed on the surface of pumice tuff in alkaline pH region. The sorption behavior of Ni and Pd in the presence of NH₃(aq)/NH₄⁺ showed considerably different trend against pH. A significant effect of co-existing NH₃(aq) on the Pd sorption behavior was confirmed and the D values were well explained using the surface complex model. This study was funded by Ministry of Economy, Trade and Industry of Japan in FY 2011. [1] A. Kitamura and T. Sasaki, Proceedings of GLOBAL 2011, Chiba, Japan (2011).

Pyrochemical reprocessing of nuclear fuels, in which electrochemical separation of actinides and fission products is mediated by a molten alkali chloride salt (typically a LiCl-KCl eutectic) is of interest for future nuclear energy cycles. A key challenge in the management of pyrochemical reprocessing wastes is decontamination and recycling of the molten salt medium to remove entrained actinides and radioactive lanthanide fission products. Since pyrochlore oxides are promising candidates for the immobilisation of lanthanides and actinides, we sought to use the “problematic” molten salt to our advantage as a reaction medium for low temperature synthesis of titanate pyrochlores. Through control of TiO₂ source, molten salt composition, reaction time and temperature, we demonstrated the synthesis of lanathanide pyrochlores at temperatures as low as 700 °C in 1 h, compared to (typically) 1350 °C in 36 h for conventional solid state synthesis. The importance of this study is in demonstrating the potential feasibility for decontamination of pyrochemical reprocessing wastes by simple addition of TiO₂ to form lanthanide and actinide pyrochlores by rapid molten salt assisted reaction at moderate temperature.

Vitrification is the currently accepted technology for the immobilisation of the high level radioactive wastes arising from the reprocessing of nuclear fuels. However, vitrification is also of increasing interest for treatment of intermediate level wastes. The purpose of the vitrified wasteform is to function as the primary barrier to the release of radionuclides to the environment, within a multi-barrier geological disposal facility. The safety case for disposal of vitrified products will require an understanding of the impact of γ-radiation (arising from the decay of fission products and actinides) on the long term physical and chemical stability of the materials. There is an absence of data concerning the performance of UK vitrified products in this regard and we have therefore performed a seminal study of the effect of γ-radiation on a variety of model UK vitrified wasteforms up to a dose of 8 MGy. It was determined that γ-irradiation up to this dose had no significant effect upon the mechanical properties of the wasteforms and there was no evidence of residual structural defects. FTIR and Raman spectroscopy showed no evidence of radiation directly affecting the silicate network of the glasses. The negligible impact of this γ-irradiation dose on the physical properties of the glass was attributed to the presence of multivalent ions, particularly Fe, and a mechanism by which the electron-hole pairs generated by γ-irradiation were annihilated by the Fe2+ - Fe3+ redox mechanism.

Vitrification is a potential treatment option for radioactive wet intermediate level wastes arising from decommissioning of UK nuclear sites. From consideration of the radiological, chemical and physical composition of wet ILW arisings from a particular UK reactor site, four waste stream blending options were identified for vitification, based on the content of inorganic metal oxides. However, the high organic and sulphur content, particularly associated with spent ion exchange (IEX) resins, of these wastes was considered potentially problematic with respect to phase separation and extremely reducing melt conditions. Taking a materials informatics approach, we compiled a bespoke database of 80 glass compositions, populated with measured and modelled property data. Applying processing and waste product acceptance criteria, defined by the waste producer and licensing authority, the database was screened for potential glass formulations meeting criteria such as: upper melting temperature; viscosity range; minimum chemical durability; acceptability criteria for Cs volatilisation; wasteform homogeneity; and waste loading. Grouping of candidates on a Pass / Fail criterion enabled identification of four suitable potential glass forming families, from which 8 target glass compositions were identified. In subsequent research, all eight compositions were demonstrated to outperform the specified criteria. The importance of this work is in demonstrating the utility of materials informatics in order to rapidly, efficiently, and successfully prototype bespoke wasteforms for radioactive waste immobilisation.

Glasses in the Al2O3-B2O3-Fe2O3-Na2O-SiO2 System were produced at a temperature of 1150 C, poured onto a metal plate and annealed. The nature of the structural units and their bonding in the structure were studied by infrared and Raman spectroscopic techniques. Structural network of all the glasses studied is built from major [SiO4] tetrahedra with 2-3 non-bridging oxygens (NBO). Incorporation of Fe2O3 offers destructive effect in glass network.

Valence state and local environment of Fe in complex glasses related to the system Al2O3-B2O3-Fe2O3-Na2O-SiO2 were studied. In all the glasses major fraction of Fe exists as Fe3+ ions but minor Fe2+ ions especially in the glass with the lowest K=[SiO2]/[B2O3] ratio is also present. Average Fe-O distance in the first shell is 1.80-1.85 Å and coordination number is 4-6. Appearance of the second sphere is rather weak demonstrating homogeneous distribution of Fe ions.

Yellow-Phase rarely occurs during vitrification process for high-level radioactive waste glass (HLWG). Major components of Yellow-Phase are molybdates which have high water solubility. Yellow-Phase includes high concentration of radioactive elements such as Cs, and Tc compared to the glass matrix. Due to these features, these radioactive nuclides could pass through canister and touch buffer materials at high concentration when Yellow-Phase exists in HLWG. Therefore, it&’s important for safety assessment of HLWG disposal to accumulate the knowledge about migration behavior of high concentration nuclides in compacted bentonite. In this study, kunipia-F was used as bentonite specimen and Cs₂MoO₄ was used as tracer. Ten micro liter of tracer solution containing 0.1M, 1M or 5M of Cs₂MoO₄ or several milligram of solid Cs₂MoO₄ was spiked on the interface between an acrylic resin coupon and compacted bentonite which dry density was 0.8 to 1.4 Mg/m³. After diffusion periods, bentonite specimen was sliced in step of 0.5 to 2mm, and amount of cesium and molybdenum in each slice was measured with ICP-MS. From the results using 0.1M of Cs₂MoO₄ solution, logarithmic concentration of cesium were proportional to squared distance from the interface, and apparent diffusion coefficient could be obtained. Apparent diffusion coefficients obtained for cesium were 2 to 7 x 10^-12 m²/s and similar to previous values and were independent of diffusion periods. On the other hand, broader concentration profiles of cesium were observed in the case of using 1 or 5M of Cs₂MoO₄ solution or solid compared with the case of using 0.1M of Cs₂MoO₄ solution. Besides that, cesium tended to diffuse faster as diffusion period was longer. Apparent diffusion coefficients in the case of using 1 or 5M of Cs₂MoO₄ solution or solid were estimated a few times bigger than those in the case of using 0.1M of Cs₂MoO₄ solution. Thus it was suggested that apparent diffusion coefficient of cesium could depend on tracer concentration. For molybdenum as chemical analog of Tc, apparent diffusion coefficients were obtained 6 to 25 x 10^-12 m²/s from the results using 0.1M of Cs₂MoO₄ solution. When 0.1M of Cs₂MoO₄ solution was used, much smaller amount of molybdenum compared to cesium could diffuse into bentonite. On the other hand, amount of molybdenum was half or more than those of cesium when 1 or 5M of Cs₂MoO₄ solution or solid was used. From these results, it was suggested that molybdate ion was much affected by anion exclusion when 0.1M of Cs₂MoO₄ solution was used, however the effect of anion exclusion decreased with increasing ionic strength.

Two explosions at the Fukushima Daiichi nuclear power plant on 12th and 14th March 2011 caused an uncontrolled release of radioactive nuclides into the environment. Dose rates of the released radioactive nuclides that made landfall on mainland Japan are largely controlled by the ground level contamination and accumulation of Cesium137 (Cs137) in populated areas. Elevated dose rates were recorded after the realease, but which have continued to decrease daily. Dose rates greater than 2.8 µSv/h were recorded in some parts of Fukushima-city in May 2011, located some 60 km NW of the Fukushima Daiichi plant, were considered to be unacceptable. The Ministry of Education, Culture, Sports, Science and Technology-Japan approached the Japan Atomic Energy Agency (JAEA) to develop an immediate and effective method of reducing the dose rate received by students in Fukushima-city&’s school facilities. Selectively and rapidly removing Cs 137 was not a pragmatic solution under the requested criteria. An effective means of reducing dose rate was to remove the top 5 cm of soil and to bury it on-site under fresh or deeper excavated uncontaminated soil for shielding. A demonstration of this method was carried out by JAEA at a junior high school yard and kindergarten playground in the center of Fukushima-city. From this demonstration it was found that: 1. Distributions of radioactive nuclides and dose rate were almost uniform throughout the junior high school yard and kindergarten playground in Fukushima-city. 2. A storage method was found without moving the contaminated soil off-site until an agreement over the final method is accepted. 3. The dose rate at ground level was reduced from 3.1 µSv/h to 0.16 µSv/h by the removal of the top 5 cm of soil. The dose rate at 1 m above ground level was reduced from 2.5 µSv/h to 0.15 µSv/h. 4. The residual dose rate was considered to be dominated by radioactive nuclides from adjacent contaminated objects such as trees and buildings. 5. A spike in dose rate from the actual buried storage site could not be detected.

A serious accident occurred at TEPCO Fukushima Dai-ichi Nuclear Power Plant (NPP) by the 2011 off the Pacific coast of Tohoku Earthquake on 11 March, 2011. Part of the radionuclides in atomic reactors was released and accumulated on soil surface and forest etc. over the wide range of area in Fukushima Prefecture. This work was carried out as one of the researches relating to distribution maps of radiation dose rate and soil contaminated by radionuclides etc. which Ministry of Education, Culture, Sports, Science and Technology (MEXT) promotes. Investigation on the depth distribution of radionuclides in soil was carried out as of 3 months after the accident and data relating to migration and retardation of radionuclides were obtained. In addition, the inventories of radionuclides accumulated on soil surface immediately after the occurrence of accident were estimated based on the measured results of depth distributions. The investigation using Geoslicer was conducted at totally 11 points in Nihonmatsu city, Kawamata town and Namie town, and plate-shaped soil samples of depth 50 cm to 1 m were taken. Gamma-ray decay nuclides were analysed with a Ge semiconductor detector. Both of Cs-134 and Cs-137 were detected in all investigated points, and Te-129m and Ag-110m were detected only in areas where spatial dose rates and the inventories of radionuclides have been evaluated to be high. At many points, the depth distributions of radionuclides were within 5 cm from surface except for soil at locations that are supposed to have been used as farmland. Radionuclides tended to distribute to deeper positions in soil at locations that are supposed to have been used as farmland than in soil in the surface layer. However, almost all radionuclides in soil at locations that are supposed to have been used as farmland also distributed within a depth of around 14 cm. The apparent diffusion coefficients (Da) of radionuclides derived from penetration profiles near the surface layer showed a tendency to be higher in soil at locations that are supposed to have been used as farmland (Da=0.1~1.5E-10 m2/s) than in soil in the surface layer (Da=0.65~4.4E-11 m2/s), and many Da values were nearly 1E-11 m2/s. Distribution coefficients (Kd) for Cs and I onto soil obtained by a batch method, in the range of Kd=2,000~61,000 ml/g for Cs and Kd=0.5~140 ml/g for I, were largely different between cation (Cs) and anion (I). Although Kd values are different between cation and anion, Da values for all detected radionuclides (Cs-134, 137 (cation), Te-129m (anion), Ag-110m (cation)) were similar levels. This is considered to be due to that those Da values were controlled by dispersion caused by advection by rain water. The estimated inventories of radionuclides accumulated on soil surface immediately after the occurrence of accident were consistent with the results of airborne monitoring and soil monitoring which have been released by MEXT etc.

The sensitive analysis of radionuclide migration for the scenarios on deterioration or loss of safety functions expected in HLW disposal system due to the human error (initial defective scenarios) is performed in this study. The initial defective scenario for vitrified waste is that the crack of vitrified waste increases due to external force caused by falling in handling process, and resulting enhanced glass dissolution rate. Another scenario assumes that the formation of molybdenum oxides or molybdates, known as yellow phase, results from the human error of fluid adjustment in vitrified-waste manufacturing. High glass dissolution rate in yellow phase and the formation of oxidizing condition due to molybdenum oxides are considered. In the initial defective scenario of overpack, the human errors such as transformation of overpack caused by falling in handling process, remaining stress occurred during welding operation and so on have an effect on the corrosion progress of overpack. It is assumed that the function of no infiltration of groundwater inside overpack loses immediately after the post-closure. Early high temperature in the engineered barrier leads to increase of glass dissolution rate. In two types of the initial defective scenario of bentonite buffer, the heterogeneous decrease of density in buffer result from some kinds of human errors in manufacturing or construction work of the buffer, humidity management and so on. The deterioration of buffer properties leads to the loss of restraint of advection. The advection transport decreases silica concentration in the porewater, and resulting increased glass dissolution rate. In another scenario, the loss of restraint of colloidal migration in buffer is due to the heterogeneous decrease of density. In the initial defective scenario of plugs, the loss of restraint with plugs from migration through dominant pathway in access tunnels and their vicinity results from the human errors. It is assumed that the pathway of radionuclide migration is short-circuit through the excavation disturbed zone under high permeability. Release rates for Cs-135 and Se-79 are estimated from Monte Carlo-based analysis. Maximum release rates of Se-79 and Cs-135 from natural barrier in three initial defective scenarios for vitrified waste and overpack are approximately equivalent to that in normal scenario on all safety function working. Maximum release rate of Se-79 in initial defective scenario of buffer under the condition of colloidal migration is about 30 times as high as that in normal scenario. Maximum release rate of Cs-135 in initial defective scenario of plugs is about two orders of magnitude higher than that in normal scenario. These results especially indicate the need to understand the feasibility on two types of initial defective scenario, leading to the loss of restraint for colloidal migration in buffer and the loss of restraint with plugs from short-circuit migration.

The organization of the fuel cycle is a legacy of World War II and the cold war. Fuel cycle facilities were developed and deployed without consideration of the waste management implications. This led to the fuel cycle model of an isolated single-purpose geological repository site for disposal of wastes. The MIT Future of the Nuclear Fuel Cycle study and the Blue Ribbon Commission on America&’s Nuclear Future recommended integration of the fuel cycle with the repository because of potential economic, safety, repository performance, nonproliferation, and institutional incentives. Collocation and integration of reprocessing and other backend facilities with the repository alters the functional requirements for waste forms and broadens the choice of waste forms. It eliminates transportation constraints and relaxes the incentives to minimize waste volumes. Waste forms can be chosen based only on repository performance and costs. This enables the use of a much wider set of waste forms. Beyond the particular chemical form, there are other implications. 1. Safeguards termination. If wastes contain significant quantities of plutonium and other fissile materials, there is a requirement for multi-generational long-term repository safeguards. Dilution of such wastes can make the fissile materials “not practically recoverable” and safeguards can be terminated before disposal based on IAEA limits. For closed fuel cycles, co-siting reduces waste volume constraints to make this a potentially practical option where maximum waste loadings would be determined by safeguards. 2. Solubility-limited waste forms. The release rates of some troublesome radionuclides (Carbon 14, iodine) from a repository are limited by the solubility of the specific radionuclide in groundwater. If the specific radionuclide is diluted by a factor of a thousand with the non-radioactive isotopes of that element, its concentration in groundwater is reduced by a factor of a thousand which should lead to a commensurate reduction in radionuclide release rates to the environment. Isotopic dilution is the most direct way to improve repository performance for solubility-limited radionuclides. 3. Limiting radiation damage. Waste forms with high concentrations of radionuclides are degraded by (1) long term radiation damage to the waste form and (2) change in the chemical composition of the wastes caused by the decay of radionuclides into different elements. Both effects are reduced by using waste forms with lower waste loadings. The paper discusses what functional requirements may change, why, what waste forms are most impacted, and the implications for future waste forms choices--including some examples. 1. C. W. FORSBERG, “Coupling the Backend of Fuel Cycles with Repositories”, Nuclear Technology, (November 2012). 2. M. KAZIMI, E. Moniz, C. W. Forsberg, et. al., The Future of the Nuclear Fuel Cycle, Massachusetts Institute of Technology (2011).

An overview of selection criteria for waste processing and disposal technologies is given. A systematic approach for selection of an optimal technology is proposed. Optimal selection of a technical processing and disposal option is case specific to the waste management needs. Waste streams considered are from nuclear applications, research, power generation, nuclear fuel cycle activities and decommissioning of nuclear facilities as well as for NORM-containing waste. Technical options and technologies are crucial for safe management of radioactive waste. Selection among available options and technologies can be done on a national level either by waste generators or by waste management organizations. The selection principals may vary from organizational preference, collected or known experience or following an optimization procedure. In any case, because of the costs involved, the potential complexity of technical and environmental considerations, and the necessity to assure adequate performance, the selection mechanism will always require rather clear criteria to address waste management needs. Some criteria will be fairly general and applicable to almost any waste management system. Others may apply to specific waste categories or to particular waste management steps. To address complex waste management needs it is essential to analyze the waste generation and understand properties, types and volumes of waste before selection of a particular technology. Furthermore it is necessary to comply with the appropriate regulatory regime, and to consider disposal options available assuming that legal and regulatory infrastructure exists or is going to be established. The selection of a technology needs to be based on the evaluation of all relevant criteria and constraints. Techniques such as multi-attribute analyses that should consider all the relevant criteria, constraints and conditions, their interactions and weights can be used to select the appropriate technology. There were few attempts to use a methodological approach on selecting an optimal technological approach. It is expedient to have a systematic approach on selecting optimal technical solutions for processing and disposal of radioactive waste which is the objective of this paper.

The evaluation of the redox conditions in the Swedish intermediate and low level waste (ILW-LLW) repository, SFR-1, is of high relevance in the performance assessment. SFR-1 repository contains heterogeneous types of wastes, of different activity levels and with different materials in the waste and in the matrices and packaging. Steel and concrete-based materials are ubiquitous in the repository. The assessment presented in this work is based on the evaluation of the redox conditions and of the reducing capacity in 12 individual waste types (representative of the different waste package types present or planned to be deposited in the SFR-1). A combination of the individual models is used to assess the redox evolution of the different vaults in the repository. The results of the model indicate that in the initial time after repository closure, oxygen is consumed through degradation of organic matter and metal corrosion. Afterwards, the system is kept under reducing conditions for long time periods, and hydrogen is generated due to the anoxic corrosion of steel, giving rise to the production of magnetite as main corrosion product. The time at which steel is exhausted varies with the amount and characteristics of steel in the different parts of the repository, and ranges from 5,000 years to more than 60,000 years. After complete steel corrosion, the reducing capacity of the system is mainly given by magnetite. The redox potential imposed by the anoxic corrosion of steel and hydrogen production is in the order of -0.75 V at pH 12.5. In case of assuming that the system responds to the Fe(III)/Magnetite system, and considering the uncertainty in the pH due to the degradation of the concrete barriers, redox potentials the range -0.7 to -0.01V are calculated. Probabilistic sensitivity analyses on the steel corrosion rates show the impact on the time at which steel is completely corroded. Simulations assuming presence of oxic water due to glacial melting, intruding the system at 60,000 years after repository closure, indicate that magnetite is progressively oxidising and forming Fe(III) oxides. The time at which magnetite is completely oxidised varies depending on the amount of steel initially present in the waste package.

The French Planning Act on radioactive waste management voted in 2006 indicates that, in order to commission of a geological repository by 2025, a license application must be submitted and reviewed by the competent authorities by 2015 for the industrial project of this geological repository called Cigéo (Centre Industriel de Stockage Géologique). On the basis of its preliminary design set up in 2009 and on the associated requirements for long-term safety, a global conceptual model has been developed in order to prepare the performance and safety analysis to be conducted as soon as 2013 and to help the dimensioning of architectures to be properly done and new developments and optimization to be introduced in the application from the current studies conducted until 2014. After its closure, the fundamental objective of a radioactive waste repository is to protect humans and the environment from the risks associated with the release of radioactive and chemical toxics. To meet these objectives, the principle of the Cigéo repository is to passively respond to different functions: - isolate the wastes from surface phenomena and human intrusions; - limit the transfer of radionuclides and chemical toxics to the biosphere. This implies (i) to limit as much as possible the presence of these chemicals, likely to contribute to the impacts, in engineered materials of the repository structures (containers, components) (ii) control the physico-chemical degradation of waste packages and containers to verify the containment of most of the radionuclides and chemical toxics close to the waste packages (iii) to identify and quantify the potential migration pathways. Such functions rely on the performance of the both engineered and geological barriers which allow: - to oppose water circulation, with repository designs favouring the limitation of the water flows and therefore of the convective transfer of solutes; - to limit the release of radionuclides and chemical toxics; - to delay and mitigate the migration of radionuclides and chemical toxics potentially released from the waste packages. In order to evaluate those performances, a conceptual model of the thermo-hydro-chemico-mechanical evolution of the different components of the repository has been designed. It takes stock of a 20 years research process which allowed data to be obtained from surface geological campaigns, in situ experiments in URLs and wastes characterization, and numerical simulation progresses to be promoted. Based on the best available knowledge, this conceptual model constitutes a robust basis for the definition of the long-term safety scenarios. It also helps identifying the residual uncertainties and provides guidelines for additional research and optimizations to be conducted.

Sulfate-containing wastes pose problems of limited solubility and retention in alkali borosilicate glasses. The past studies to optimize the composition of a glass frit for combination with the waste to improve sulfate retention while meeting other process and product performance constraints have resulted in a specific frit composition and a concentration limit for sulfate in the glass recommended for the next batch of sludge to be processed at Savannah River Site (SRS). Additionally, the sulfates can accumulate at the top of the melter, corroding the melt-contact materials. A maximum concentration of about 1.5 wt% has been reported in these glasses. We have studied sulfate solubility and retention in simulated borosilicate nuclear waste glasses. Our main objective is to increase waste loading of sulfur-containing wastes. We have selected the baseline borosilicate glass system and the simulated nuclear waste components developed by the Savanna River National Laboratory (SRNL) for our study. Current target sulfur level is 0.6 wt% at SRS, while our new target levels are 0.8 and 1.0 weight%. We have added sulfur in the form of sodium sulfate normalized into the glass network to our target levels. We have selected minor additives, barium, lead, and vanadium, to the predetermined glass matrix to increase the sulfate solubility. The solubility of these glasses will be determined initially by visual observation and then using the Inductively Coupled Plasma Atomic Emission Spectroscopy (ICP-AES). We have also studied the sulfate volatilization of these glasses. We have used ICP-AES to determine the composition of the glasses. We have used well-established characterization tools (e.g.., UV-Vis-NIR, Raman, and Fourier transform infrared (FTIR) spectroscopy) to study the glass structure. We have also used thermal analytical tools (differential thermal analysis (DTA), thermogravimetric analysis (TGA), and differential scanning calorimetry (DSC) to study the thermal stability of the glasses. Normalized solubilities and volatilization rates will be reported.

The resulting wastes from spent fuel reprocessing (fission products and minor actinides) are currently immobilized into a borosilicate glass known as R7T7. Its long-term behavior has been investigated with a view to geological disposal, and the kinetics and mechanisms of glass alteration have been studied in aqueous media. Most of the current knowledge of the long term behavior is based on studies of non-radioactive material. These investigations have conducted to the observation of several stages during glass alteration: the first corresponds to a congruent dissolution characterized by the initial dissolution rate. Then in-situ recondensation of dissolved species results in the formation of a gel that decreases the alteration rate by several orders of magnitude to a residual alteration rate. The present work focuses on the study of the effects of electronic dose rate on this residual alteration rate by studying both doped glasses and glasses submitted to external gamma irradiation. A 0.59%wt 239Pu doped glass (α emitter) and a 0.16%wt 99Tc doped glass (β emitter) are studied here under static conditions. These glasses are leached under argon atmosphere, at 90 °C and at a high surface-area-to-volume ratio (SA/V = 30 cm-1), in order to reach quickly the residual alteration rate. In addition, leachings of non-radioactive glasses under external irradiations are carried out in a gamma irradiator with 60Co sources. Three dose rates are studied under Ar atmosphere, at 90°C and with SA/V = 100 cm-1: two very high dose rates of 5 and 10 kGy.h-1 in order to exacerbate radiation effects, one lower dose rate (0.05 kGy.h-1), in order to compare with experimental results on doped glasses. The alteration rate is monitored by the releases of glass alteration tracer elements (B, Na and Li), which are measured by ICP-AES. Radiation effects on the glass leached and its gel network are characterised by Scanning Electron Microscopy (SEM) and Transmission Electron Microscopy (TEM). Plutonium and technecium releases are also measured by radiometry and their chemical oxidation state is assessed by measuring both pH and redox potential of the leachates. Results do not underline any significant effect of irradiation on the residual alteration rate of doped glasses. Similar results have been observed on the leached glasses under external gamma irradiation up to 10 kGy.h-1. These observations are consistent with SEM and TEM characterizations which show that a protective layer can be formed under irradiation. Concerning the behavior of plutonium and technetium, very low concentrations (around 10-7 - 10-8 mol.L-1) of these elements are measured in the leachates, traducing their high retention in the alteration layer and/or sorption on the walls of the reactor.

All nuclear processes produce radioactive waste and, to prevent the release of radionuclides into the biosphere, many countries immobilize the waste into durable waste forms. Demonstrating their long-term performance is difficult, however. Many glasses are extremely durable, with manmade glasses intact after thousands of years in aqueous environments and natural obsidians lasting for millions of years. Considering this, the minimal reaction progress observed at laboratory time scales is nearly irrelevant; small errors in short-term measurements would lead to completely unacceptable error ranges when simply extrapolated to geologic time-scales. Therefore, a robust scientific and mechanistic understanding of the processes responsible for glass degradation and radionuclide release is necessary to develop technically defensible models that can be used to calculate the long-term behavior of a glass waste form and its ability to control the release of radionuclides to the ground water at some reasonable margin of safety below that required in the applicable regulations. Radionuclide release from glass is due to is the combined contributions of mechanisms that can be placed into two general categories: reaction control (impacted by solution affinity, local solution changes due to secondary phase precipitation, and kinetic limitations) and transport control (impacted by ion exchange, pore tortuosity, and diffusion of rate-limiting ions). One key remaining question is the mechanism and extent that alteration layers present a barrier to corrosion. A detailed understanding of the location, composition, evolution, and function of transport-limiting layer(s) would enable the accurate modeling of the rate-limiting mechanism of radionuclide release. Isotopic substitution provides a mechanism for tracing the progression of ions through mature alteration layers while not changing the system chemistry. Compositionally equivalent glasses with different isotopic ratios of several elements were synthesized and allowed to corrode in equivalent static conditions for just over one year. Following the development of a mature alteration layer, the solutions were swapped. The concentrations of individual isotopes were monitored both in solution through ICP-MS and alteration layer depth profiles through secondary ion mass spectroscopy and atom probe tomography. This experiment is the first to simultaneously target glass formers (Si, B), an alkali modifier (Li), an alkaline earth modifier (Ca), and several transition metals (Fe, Zr, and Mo). The results show that several previous assumptions regarding ion transport in corroding glass systems are incomplete. Substantial and rapid participation of several monitored ions in solution with the innermost alteration layers was observed. Lithium exhibited diffusion far into the glass in remarkably short time-scales. Based on these results, new mechanistic models of glass corrosion are proposed.

The development of iodine immobilization technique that can contain iodine within the waste form for long periods of time and constrain its leaching into pore water is desired in order to secure long term safety of geological disposal of TRU waste. Lead borate glass vitrified by low temperature is regarded as one of the promising immobilization techniques to contain Iodine-129 which will be removed from spent AgI filter generated from reprocessing plants in Japan and may cause significant effect onto long term safety of geological disposal. Batch corrosion tests were conducted to understand glass dissolution behavior in various solutions that simulated geological disposal conditions. As a result, boron dissolved with the highest rate in all types of solution to be identified as an index element which could represent glass dissolution rate. On the other hand, lead dissolved with much slower rate. These results of batch corrosion tests were consistent with EPMA observations of altered glass surfaces, that boron was depleted in altered layer but lead was rich on the contrary. In addition, altered glass surfaces were observed by SEM, TEM and XRD. SEM observation suggested that altered surface layer was composed by 2 different types of layers: one was rather porous media next to pristine glass, another was aggregates of trigonal crystals with size of some micrometers. XRD analysis showed that the trigonal crystal was identified as hydrocerussite (Pb3 (CO3)2(OH)2), that could be predicted by geochemical simulation as well. 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).

At Sellafield, the Post Operational Clean Out (POCO) of solids from the base of the highly active waste storage tanks, in preparation for decommissioning, will result in a high molybdenum stream which will be vitrified using the current Waste Vitrification Plant (WVP). In order to minimise the number of containers required for POCO, the high molybdenum feed could be co-vitrified by addition to reprocessing waste, using the borosilicate glass formulation currently utilised on WVP. Co-vitrification of high molybdenum feeds has been carried out using non-active simulants, both in the laboratory and on the Vitrification Test Rig (VTR) which is a full scale working replica of a WVP processing line. In addition, a new borosilicate glass formulation containing calcium has been developed by NNL which allows a higher incorporation of molybdenum through the formation of a durable CaMoO₄ phase, after the solubility limit of molybdenum in the glass has been reached. Vitrification of the high molybdenum feed in the presence of varying quantities of reprocessing waste liquor using the new glass formulation has been carried out in the laboratory. Up to ~10 wt% MoO₃ could be incorporated without any detrimental phase separation in the product glass, but increasing the fraction of reprocessing waste was found to decrease the MoO₃ incorporation. Soxhlet and static powder leach tests have been performed to assess the durability of the glass products. This paper discusses the results of the vitrification of high molybdenum feeds in the presence of reprocessing liquor in both the borosilicate glass formulation currently utilised on WVP and the modified formulation which contains calcium.

Glasses in the Al2O3-B2O3-Fe2O3-Na2O-SiO2 System were produced at a temperature of 1150 C, annealed, and examined using XRD and SEM/EDX. Surfaces of same samples were additionally heat-treated and etched with HCl. The pristine samples were X-ray amorphous and rather homogeneous except B1 sample contained trace of carnegieite/nepheline and spinel. Corrosion of these glasses under etching proceeds by conventional mechanism with a damage of their surface layers but the drop-type structure was observed after etching of the surface of glass B2 that suggests occurrence of liquid-liquid phase separation.

Under deep geological repository conditions, there are two different processes of release of radionuclides i) the Instant Release Fraction (IRF) which refers to the fraction of radionuclides rapidly released and it depends on the radionuclides concentration transported to spent fuel rod areas with low confinement, such as gap and grain boundaries and ii) the long term release of radionuclides confined in the spent fuel matrix. In terms of the long term release of radionuclide, the Matrix Alteration Model (MAM) is one of the most evolved models describing the dissolution mechanism in which an Alteration/Dissolution source term model is based on the oxidative dissolution of spent fuel. Oxidant and reducing species can be naturally or radiolytically-generated. The experimentally-observed inhibition of matrix dissolution by H2 is rationalized by activation of hydrogen through two main pathways: a) the alfa, beta and gamma radiation through reactions with radiolytic radicals and b) by the metallic particles contained in the fuel surface (epsilon particles). The activation of hydrogen by radiation was integrated into MAM with the reaction: OH+H2=H2O+H and validated with experimental data. The effect of ε particles on H2 activation has been introduced in MAM, the proposed mechanism is in agreement with a galvanic coupling between spent fuel matrix and epsilon particles. The mechanism consist of a first step controlled by diffusion of H2 on the ε particle surface, which determines the kinetics of the process, and a rapid step of electrons being transferred from the epsilon particle to the matrix. Under deep repository conditions and at the time expected of water contacting the spent fuel (after 1000 years of storage), α radiation will represent the main contribution to water radiolysis. In the current study, the calculations are performed considering different particle sizes of spent fuel and evaluating the effect of surface area on the alteration/dissolution of spent fuel matrix.

Actinide silicates such as USiO₄ coffinite are expected to play an important role in the field of direct storage of spent nuclear fuels in underground repository since they could control the concentration of actinide in solution. In this framework, this study aimed to investigate the formation of actinide-bearing silicates and collect associated thermodynamics data, including solubility constants. A series of Th_1-xU_xSiO₄ samples was thus prepared by hydrothermal method (250° C, autogenous pressure) then characterized by XRD, mu;-Raman spectroscopy and ESEM. XRD analysis systematically evidenced the formation of pure Th_1-xU_xSiO₄ samples for 0 le; x le; 0.2 while the formation of additional mixed oxide (Th_1-yU_yO₂) was always detected for higher substitution rates. This latter was found to be nanosized and usually presented a poor crystallinity. In order to determine the kinetics leading to the formation of pure silicate based solid solutions, the evolution of the Th_0.5U_0.5SiO₄ samples was explored for various heating times. The results evidenced an increase in the amount of silicate with the heating time, which could indicate a slow-rate oxide to silicate phase-transition under hydrothermal conditions. Such slow kinetics then appeared to preclude the preparation of pure phases with high uranium incorporation rates, typically x ge; 0.8, at the laboratory time-scale. In these conditions, purified Th_1-xU_xSiO₄ samples were further obtained by the means of either physical or chemical separation methods. The refinement of the corresponding lattice parameters in the zircon-type I41/amd space group finally confirmed the formation of a complete solid solution between ThSiO₄ thorite and USiO₄ coffinite. In a second step, the dissolution kinetics of uranothorite samples were assessed through both static and dynamic leaching tests. As an example, the normalized dissolution rates varied from 2.10^-6 to 4.10^-9 g.m^-2.day^-1 depending on the composition considered (10^-1M HCl, room temperature). Also, when reaching the thermodynamic equilibrium, first values of the solubility constants at room temperature were evaluated for 0.1 le; x le; 0.5. In this aim, the activity of aqueous species involved in the precipitation reaction were calculated thanks to the PhreeqC software and led to K_S values ranging between 1.10^-3 and 8.10^-3 which appeared slightly higher than that usually reported in the literature for coffinite.

As an analog for behavior of spent nuclear fuel, the corrosion behavior of simulated nuclear fuel (SIMFUEL) in a high level radioactive waste repository disposal setting was investigated using electrochemical and solution chemistry analyses. The SIMFUEL was exposed to aerated solutions of NaCl+NaHCO3 with and without calcium and silicate. Two SIMFUEL compositions were studied, representing spent nuclear fuel (SNF) with 3 or 6 (atomic) percent burnup in terms of fission products equivalents of surrogate elements. For all tested cases, the corrosion potential of the SIMFUEL increased with time and reached steady state values in the range of 0.06 to 0.12 VSCE within 72 hours. The polarization resistance increased with increased immersion time, indicating possible blocking effects due to accumulation of corrosion products on the SIMFUEL electrode. The potential-pH diagram suggests formation of schoepite, and this phase may cause the increase in the polarization resistance with time. The addition of calcium and silicate produced no measureable change in values of the polarization resistance measured at the corrosion potential. The impedance was similar for solutions with or without calcium and silicate, indicating a relatively minor effect of calcium and silicate on the oxidation and dissolution rate of SIMFUEL. The dissolution rate, estimated by applying both the Stern-Geary Equation and Faraday&’s Law, ranged from 1 to 3 mg/m2-day, which is similar to the range of dissolution rates for SIMFUEL and SNF reported in the literature for comparable conditions. SIMFUEL with simulated high burnup (6 percent) resulted in a minimally higher dissolution rate than the samples with simulated lower burnup (3 percent). Analysis of the solution chemistry shows that uranium is the dominant element dissolved in the post-test solutions, and the dissolution rates calculated from uranium concentration are consistent with the dissolution rates obtained from impedance measurements. Simulated-fission product elements (i.e., barium, molybdenum, strontium, and zirconium) dissolved from the SIMFUEL electrode at a relatively high rate. Sorption test results showed that uranium concentrations in the post-test solutions with a stainless steel disk decreased about 20 percent after 21 days of immersion of the stainless steel compared to the concentrations without immersed stainless steel, suggesting sorption of uranium onto the stainless steel oxide. Electrochemical impedance was found to be an effective technique for measuring uranium dissolution rate in real time. SIMFUEL characterization results indicate that SIMFUEL is a valid analog material to evaluate the SNF matrix dissolution rate after long-term containment, when gamma and beta radiation have decreased significantly in a repository environment. This abstract is an independent product of the CNWRA and does not necessarily reflect the view or regulatory position of NRC. The NRC staff views expressed herein are preliminary and do not constitute a final judgment or determination of the matters addressed or of any licensing action that may be under consideration at NRC.

The release of radionuclides from spent nuclear fuel in contact with water is controlled by two processes - the dissolution of the UO2 grains and the rapid release of fission products segregated either to the gap between the fuel and the cladding or to the UO2 grain boundaries. The rapid release is often referred to as the instant release fraction (IRF) and is of interest for the safety assessment of geological repositories for spent fuel due to the potential dose contribution. Previous studies have shown that the instant release fraction can be correlated to the fission gas release (FGR) from the spent fuel. Studies comparing results from samples in the form of pellets, fragments, powders and a fuel rodlet have shown that the sample preparation has a significant impact on the instant release, indicating that the differentiation between gap release and grain boundary release should be further explored. There is a trend for many reactors towards increasing fuel burnup. Furthermore, fuels with additives and dopants in the fuel matrix are more commonly used. The additives and dopants affect the fuel properties such as grain size and fission gas release. In the present study we have performed studies using different high burnup fuels with and without additives/dopants and compared the fuel types with respect to their instant release behavior.

The Instant Release Fraction (IRF) concept can be defined as all radionuclide fractions that will be released during a short period of time when water reaches the fuel. A theoretical definition of IRF would comprise the total inventory of radionuclides segregated from the UO2 matrix and located in the gap between the fuel and cladding and at the grain boundaries. The radionuclide release from the high burn-up structure has recently determined to be much slower than expected and it would not be included in the IRF. Also, not all the segregated radionuclides should be taken into the IRF since some of them are anyway going to be dissolved much slower than the spent nuclear fuel matrix. In order to study the influence of burn-up as well as gap and grain boundary contribution on IRF, static leaching experiments were performed with four commercial UO2 SNF: three from a Pressurised Water Reactor (PWR) with a BU of 48, 52 and 60 MWd/kgU and one from Boiling Water Reactor (BWR) with a BU of 53 MWd/kgU. All the experiments were performed at room temperature (24 ± 6) C under oxic conditions by using bicarbonate water (1mM HCO3-) as a leachant. Results have shown that the elements that constituted the IRF in all the experiments were: Rb, Sr, and Cs and in a conservative approach Mo and Tc. The IRF (%) values were calculated and compared showing IRFgap > IRFgrain boundaries for Cs and Rb, which would indicate a greater segregation into the gap between the cladding and the pellet and fractures than into the grain boundaries. On the contrary, Sr, Tc and Mo showed IRFgap le; IRFgrain boundaries which would indicate that these elements are found in the grain boundaries in a higher proportion than Rb and Cs. A semi-empirical model has been developed in order to explain IRF taking the allocation of the different radionuclides as well as the contact with water into account.

In the safety case for the geological disposal of nuclear waste, the release of radioactivity from the repository is controlled by the dissolution of the spent fuel in groundwater. Therefore, to assess the performance of the repository after infiltration of groundwater and contact with spent fuel, the dissolution characteristics must be determined. In spent nuclear fuel, high energy sites occur at grain boundaries and within the material as naturally occurring surface defects. Current studies of spent nuclear fuel dissolution have not considered the effect of high energy surface sites within the material structure. In this investigation, CeO2 analogues, which approximate as closely as possible the characteristics of fuel-grade UO2, were characterised after dissolution under a wide range of conditions. Samples were powdered to three different size fractions to investigate high energy surface site density on dissolution rates, while monolith samples were monitored for development of surface defects such as pores and dissolution pits, and step edge rounding in addition to morphological changes at grain boundaries and surface pores. The samples were subject to a range of aggressive and environmentally relevant alteration media with different solubility controls, and reacted at a range of temperatures from 25°C to 150°C. Dissolution kinetics were monitored through analysis of the coexisting aqueous solution, and morphological changes at the surface using in-situ atomic force microscopy, confocal profilometry, vertical scanning interferometry and scanning electron microscopy. Dissolution rates were found to be greatest in low pH solutions and at higher temperatures. In neutral and basic solutions, precipitation of Ce(OH)4 was observed. Dissolution increased as a function of high energy site density, with most dissolution occurring at grain boundaries and around surface pores.

A variety of characteristics of spent nuclear fuel (SNF) is considered for postclosure performance assessments for high-level nuclear waste disposal. In developing methods for evaluating radionuclide release, it is important to consider the characteristics of the waste forms that will affect release and radiological dose. Burnup of SNF is a factor that influences those characteristics. Projections from the literature indicate an upward trend in SNF burnup and enrichment levels in the United States. A variety of potentially performance-altering factors (e.g., actinide and fission product inventory, thermal output, cladding degradation rate, instant release fraction, and SNF degradation) are associated with higher SNF burnup. The objective of this study is to quantitatively assess the relative impacts of these factors on the performance of a hypothetical disposal system. Analyses are being carried out using the computer code Scoping of Options and Analyzing Risk (SOAR), a generic, non-site-specific geologic disposal system performance assessment model developed as a tool to increase the U.S. Nuclear Regulatory Commission&’s (NRC) preparedness for regulating alternative disposal systems and geologic disposal sites. SOAR is being used to risk inform staff on factors and components that may be novel and may significantly impact NRC&’s regulatory readiness. Prior to using SOAR, ORIGEN-ARP is used to compute radionuclide inventories in the representative rim region of the high burn-up fuel pellets. The SOAR model then uses the radionuclide inventory data to compare the relative effects of high-burnup SNF on radiological dose. Preliminary analyses conducted to date involved only the inventory effects on radiological risk. The analyses suggest approximately 6 percent change to the peak expected dose from the high-burnup SNF (i.e., 60 GWd/MTU; 5 percent enrichment) when compared to the lower burnup (i.e., 40 GWd/MTU; 4 percent enrichment). Other analyses being carried out involve consideration of (i) fuel rod cladding failure time; (ii) increased fission products in the early or instantaneous release fraction of the radionuclide inventory; and (iii) potential effect of microstructural changes (e.g., formation of submicron-sized grains, development of high-pressured micro-pores in the rim region, and fragmentation due to thermal cycling) on SNF dissolution rate, and (iv) potential chemical reactivity changes (e.g., catalytic effect on dissolution of some fission products). The impact of higher thermal output of the high burnup SNF resulting from higher content of actinides and long-lived fission products will be discussed only qualitatively. This abstract is an independent product of the CNWRA and does not necessarily reflect the view or regulatory position of NRC.

Presently, an Experimental Demonstration Centre (EDC) for testing innovative technologies of spent nuclear fuel processing is created at Mountain Chemical Combine in the Russian Federation. Minimization of radioactive waste is one of the main requirements to the technologies being tested. The report represents the results of research and development on cementation of liquid radioactive waste (LRW), resulting from the spent nuclear fuel reprocessed at the EDC. LRW of EDC can be conditionally divided into non-technological waste and technological waste. Non-technological LRW of EDC includes the so-called “floor drain water” - low-saline low-active solutions consisting of water from the decontamination of premises, equipment leaks, water of sanitary locks, sanitary inspection rooms, showers, etc. Technological LRW of EDC are tritium and acetate containing solutions. Optimum cementation parameters for incorporation in an inorganic matrix, based on Portland cement and slag Portland cement were tested for these two types of waste. Bentonite or clinoptilolite are proposed to be used as a sorption additive for fixing the most leached radionuclide - Cs-137 in a compound. As to tritium water, its complete localization in the compound is possible by using an additional tracing on its surface of the bitumen film and its subsequent placement in plastic and steel containers. The obtained cement compounds completely met standard requirements for mechanical strength, leaching rate, water and frost resistance. The studies of heat generation during cement compounds hardening showed that the heat of hydration depends on the type of binder, type of sorption additives and water/binder ratio and can vary widely. The lowest specific heat characterized compounds which are based on the slag Portland cement with clinoptilolite as a sorption additive. The dry residue content in the obtained cement compounds (cement compound loading capacity) with the incorporation of non-technological and technological LRW reached 12 and 15% respectively. The results of the laboratory tests allowed to make recommendations for the development of technological regimes and industrial equipment for LRW cementation at EDC.