Kevin M. Fox, Savannah River National Laboratory
Kazuya Idemitsu, "Kyushu University"
Christophe Poinssot, Commissariat a l'Energie Atomique
Neil Hyatt, The University of Sheffield
Karl Whittle, The University of Sheffield
LL3: Fukushima Daichi
Monday PM, November 26, 2012
Hynes, Level 1, Room 109
2:30 AM - *LL3.01
Technical and Non-technical Lessons Learned from the Fukushima Nuclear Plant Accident
Akira Tokuhiro 1 Massimo Bertino 2 Micah Hackett 3
1University of Idaho Idaho Falls USA2Virginia Commonwealth University Richmond USA3TerraPower Bellevue USAShow Abstract
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.
3:00 AM - *LL3.02
Decontamination Pilot Projects to Build a Knowledge Base for Fukushima Environmental Restoration
Kaname Miyahara 1
1Japan Atomic Energy Agency (JAEA) Tokyo JapanShow Abstract
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.
3:30 AM - *LL3.03
Research and Development Activities for Cleanup of the Fukushima Daiichi Nuclear Power Station
Toshiki Sasaki 1 Masashi Saito 2 Yasuaki Miyamoto 1 Hideyuki Funasaka 1
1Japan Atomic Energy Agency Tokyo Japan2Tokyo Electric Power Company Tokyo JapanShow Abstract
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.
LL4: Waste Repositories I
Monday PM, November 26, 2012
Hynes, Level 1, Room 109
4:30 AM - LL4.01
NMR Study of Interlayer and Non-interlayer Porewater in Water-saturated Bentonite and Montmorillonite
Torbjoern Carlsson 1 Arto Muurinen 1 Andrew Root 2
1VTT Technical Research Centre of Finland Espoo Finland2MagSol Helsinki FinlandShow Abstract
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.
4:45 AM - LL4.02
Long-term Corrosion of Zircaloy 4 and Zircaloy 2 by Continuous Hydrogen Measurement under Repository Condition
Tomofumi Sakuragi 1 Hideaki Miyakawa 1 Tsutomu Nishimura 2 Tsuyoshi Tateishi 3
1Radioactive Waste Management Funding and Research Center Tokyo Japan2Kobe Steel, Ltd Kobe Japan3Kobelco Research Institute, Inc. Kobe JapanShow Abstract
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.  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.  (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 × t0.55 and the absorbed hydrogen per unit surface area (mmol/m2) = 0.04 × t0.56; 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.  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).  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).  FEPC and JAEA, Second Progress Report on Research and Development for TRU Waste Disposal in Japan (2007).  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).
5:00 AM - LL4.03
Radioelement Solubilities in SR-site, the Influence of Variability and Uncertainty
Christina Greis Dahlberg 1 Patrik Sellin 1 Miriea Grivamp;#233; 2 Lara Duro 2 Kastriot Spahiu 1
1SKB Stockholm Sweden2Amphos 21 Barcelona SpainShow Abstract
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.
5:15 AM - LL4.04
Glass-iron-clay Interactions in a Radioactive Waste Geological Disposal: A Multiscale Approach
Diane Rebiscoul 1 Emilien Burger 1 Florence Bruguier 1 Nicole Godon 1 Jean-Louis Chouchan 1 Jean-Pierre Mestre 1 Pierre Frugier 1 Stephane Gin 1
1CEA Bagnols-Sur-Ceze FranceShow Abstract
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  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.
5:30 AM - LL4.05
Use of Bioapatite as a Backfill Material for Nuclear Waste Isolation
Alyssa J. Finlay 1 Amanda E Drewicz 1 Dennis O. Terry 1 David E Grandstaff 1
1Temple University Philadelphia USAShow Abstract
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.
LL1: National Programs
Monday AM, November 26, 2012
Hynes, Level 1, Room 109
10:00 AM - *LL1.01
Waste Form Research Conducted by the Department of Energy Office of Nuclear Energy
John Vienna 1 Terry Todd 2 Kimberly Gray 3 James Bresee 3
1PNNL Richland USA2Idaho National Laboratory Idaho Falls USA3U.S. Department of Energy Washington USAShow Abstract
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.
10:30 AM - *LL1.02
Towards Sustainable Nuclear Fuel Cycles
Bernard Boullis 1
1CEA/Saclay Gif sur Yvette Cedex FranceShow Abstract
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.
11:30 AM - *LL1.03
ACSEPT and ACTINET-I3: Two Projects Gathering the European Actinide Chemistry Community
Stephane Bourg 1 Andreas Geist 2 Laurent Cassayre 3 Chris Rhodes 4 Christian Ekberg 5
1CEA Bagnols/Ceze France2KIT-INE Karlsruhe Germany3CNRS Toulouse France4NNL-UK Sellafield United Kingdom5Chalmers Goteborg SwedenShow Abstract
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.
12:00 PM - LL1.04
Impact of the Actinides Recycling on the Environmental Footprint of Nuclear Energy Systems: Comparison of Open and Closed Nuclear Fuel Cycles
Christophe Poinssot 1 Bernard Boullis 2 Christine Rostaing 1
1CEA Bagnols sur Ceze France2CEA Gif-Sur-Yvette FranceShow Abstract
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.
LL2: Glass Wasteforms I
Monday AM, November 26, 2012
Hynes, Level 1, Room 109
12:15 PM - LL2.01
Increasing the Technology Readiness of Vitrification Processes for the Treatment of UK Radioactive Wastes
Neil Hyatt 1
1The University of Sheffield Sheffield United KingdomShow Abstract
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.
12:30 PM - LL2.02
The Use of High Durability Glasses for Encapsulation of High Temperature Reactor (HTR) Fuel
Paul George Heath 1 2 Neil C Hyatt 1 Martin C Stennett 1 Owen G McGann 1
1The University of Sheffield Sheffield United Kingdom2The University of Manchester Manchester United KingdomShow Abstract
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.
Kevin M. Fox, Savannah River National Laboratory
Kazuya Idemitsu, "Kyushu University"
Christophe Poinssot, Commissariat a l'Energie Atomique
Neil Hyatt, The University of Sheffield
Karl Whittle, The University of Sheffield
LL7: Halide Solutions
Tuesday PM, November 27, 2012
Hynes, Level 1, Room 109
2:30 AM - LL7.01
Development of Advanced Waste Forms for Iodine-129
Terry Garino 1 Tina M Nenoff 1 David X Rademacher 1 Patrick V Brady 1 Dorina F Sava 1 Haiqing Liu 1
1Sandia National Labs Albuquerque USAShow Abstract
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.
2:45 AM - LL7.02
Development of the Synthetic Rock Technique for the Immobilization of Iodine: Kinetics of the Alumina Matrix Dissolution under High Alkaline Conditions
Hideaki Miyakawa 1 Tomofumi Sakuragi 1 Hitoshi Owada 1 Osamu Kato 2 Kaoru Masuda 3
1Radioactive Waste Management Funding and Research Center Tokyo Japan2Kobe Steel, Ltd. Kobe Japan3Kobelco Research Institute, Inc. Kobe JapanShow Abstract
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).
3:00 AM - LL7.03
The Study on Iodine Release Behavior from Iodine-immobilized Cement Solid
Yoshiko Haruguchi 1 Shinichi Higuchi 1 Masamichi Obata 1 Tomofumi Sakuragi 2 Ryota Takahashi 2 Hitoshi Owada 2
1Toshiba Corporation Kawasaki Japan2Radioactive Waste Management Funding and Research Center Chuo-ku, Tokyo JapanShow Abstract
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).
3:15 AM - LL7.04
Towards a Silicate Matrix for the Immobilisaton of Halide-rich Wastes
Matthew Gilbert 1
1AWE Reading United KingdomShow Abstract
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 dispos