bluuebunny/crossref_metadata_2025_split huggingface dataset:
Signatures of Non-integer 5f Occupancy in Pu Systems: Magnetic Properties and Photoelectron Spectroscopy Studies
Authors: Havela Ladislav , Shick Alexander , Gouder Thomas , Wastin Franck , Rebizant Jean 
Published: April 2011

AbstractVery diverse Pu compounds exhibit strikingly universal features in their valence-band photoemission (PES) spectra. The conjecture that such features represent the 5f5 final state multiplet has been corroborated by LDA+Hubbard I calculations, meaning that the ground state has a mixed 5f5-5f6 character. Later on, more elaborated DMFT techniques (one crossing approximation, QMC) led to similar conclusions, providing quantitative explanation of such intermeate-valent situation in more details. Analogies in PES spectra of δ-Pu and other Pu systems suggest that the situation envisaged for δ-Pu is relevant for a large group of Pu compounds. Here we show that the around mean field LDA+U in conjunction with the Hubbard I approximation, which describes well the non-magnetic ground state for δ-Pu, captures in reality properties of a large group of Pu (as well as e.g. Am) compounds, reproducing correctly the onset of magnetism and size of magnetic moments.
Thermodynamic Simulation and Experimental Study of Irradiated Reactor Graphite Waste Processing with REE Oxides
Authors: Karlina Olga K. , Klimov Vsevolod L. , Pavlova Galina Yu. , Ojovan Michael I. 
Published: April 2011

AbstractThermochemical processing of reactor graphite waste is based on self-sustaining reaction 4Al + 3TiO2 + 3C = 3TiC + 2Al2O3 which chemically binds 14C from the irradiated graphite in the titanium carbide. Thermochemical processing was investigated to analyse the behaviour of rare earth elements (REE), where REE = Y, La, Ce, Nd, Sm, Eu and Gd. Both thermodynamic simulations and laboratory scale experiments were used. The REEs in the irradiated reactor graphite are formed as activation products of impurities and spread over the graphite bricks surfaces as well as arise from fission of nuclear fuel. REEs can be used also to substitute for waste actinides as well as to increase the durability of carbide-corundum ceramics relative to waste actinides.Thermodynamic calculations and X-ray diffraction analysis of ceramic specimens synthesized revealed that durable REE's aluminates with perovskite, β-alumina and garnet structures are formed by interaction of REE oxides with the Al2O3 melt during the selfpropagating reaction of ceramic formation.The porous carbide-corundum ceramics synthesized have a high hydrolytic durability, e.g. the normalised leaching rates of 137Cs, 90Sr and Nd are of the order of 10–7 – 10–8 g/(cm2·day).
Opportunities for Research in the UK for Decommissioning and Disposal
Authors: Smart Neil 
Published: April 2011

AbstractThe NDA remit as set out within the Energy Act includes – “promote, and where necessary fund, research relevant to nuclear clean up”. The NDA need to underpin delivery and / or accelerate programmes to fulfil the overall mission and technical underpinning of these activities is critical. In this paper we will present consideration of the investment required in nuclear waste Research and Development.Firstly, NDA set the requirement for nuclear sites to write down within the Life Time Plans (LTP), at a high level, the proposed technical baseline underpinning the LTP activities; furthermore we required technology gaps / opportunities in the technical baselines to be outlined in a R&D requirements section to the LTP. Criteria were established to categorise the R&D in three areas:“needs” - those development activities needed to underpin the proposed technical solutions“risks” – those activities required to reduce / eliminate key risks to the proposed technical solutions“opportunities” – innovations / changes to the technical baselinesThe purpose of production of the technical baselines and underpinning R&D requirements is to establish an auditable trail through the LTP from programme components into how the programme will be delivered.NDA believes the production of the technical baselines and R&D requirements will be of benefit to the Site License Companies (SLC) in terms of ensuring a focus on overall programme delivery and not just short term activities. Furthermore, we can ensure that investment in technology is targeted at priority areas, with common issues and requirements identified and solutions on a broader scale will be achievable.
Cement Based Encapsulation Trials for Low-Level Radioactive Effluent Containing Nitrate Salts
Authors: Sugaya Atsushi , Horiguchi Kenichi , Tanaka Kenji , Kobayashi Kentaro 
Published: April 2011

AbstractThe operation of nuclear fuel reprocessing plants generates a radioactive effluent containing nitrate salts as the major constituent. This waste must be disposed of safely and economically, and to achieve this aim the Japan Atomic Energy Agency (JAEA) is developing a cement based encapsulation method to immobilise this waste. Non-radioactive development work has been performed at both small and large-scale (up to 200 litres) to investigate the optimum cement formulation. The results from these studies demonstrate that nitrate waste that has been concentrated by evaporation of the water to a predetermined level can be successfully encapsulated up to a waste loading corresponding to 50 wt% sodium nitrate. It has been identified that high concentrations of bicarbonate ions in the effluent can have a detrimental effect on the strength of the cement encapsulated waste; however, provided this concentration is controlled, successful encapsulation of the effluent is still achieved.
Electron Irradiation and Electron Tomography Studies of Glasses and Glass Nanocomposites
Authors: Möbus G. , Yang G. , Saghi Z. , Xu X. , Hand R.J. , Pankov A. , Ojovan M.I. 
Published: April 2011

AbstractCharacterization of glasses and glass nanocomposites using modern transmission electron microscopy techniques is demonstrated. Techniques used include: (i) high-angle-annular dark field STEM for imaging of nanocomposites, (ii) electron tomography for 3D reconstruction and quantification of nanoparticle volume fractions, and (iii) fine structure electron energy loss spectroscopy for evaluation of boron coordination. Precipitation of CeO2nanoparticles in borosilicate glasses is examined as a function of glass composition and redox partner elements. A large increase in the solubility of Ce is found for compositions where Ce retains +IV valence in the glass. Irradiation experiments with electrons and λ-rays are summarized and the degree of damage is compared by using changes in the boron K-edge fine structure, which allows the gradual transition from BO4to BO3coordination to be followed.
Glass Composite Materials for Nuclear and Hazardous Waste Immobilisation
Authors: Ojovan Michael I. , Juoi Jariah M. , Boccaccini Aldo R. , Lee William E. 
Published: April 2011

AbstractGlass composite materials (GCM) are versatile wasteforms for immobilising various types of both radioactive and hazardous wastes. We review current research on the utilisation of GCMs for hazardous and radioactive waste immobilisation. Compared to homogeneous glassy materials GCMs can incorporate larger amounts of waste elements and, in the case of glass matrix composites, they can be produced using lower processing temperatures (by viscous flow sintering) than those of conventional melting.
Waste Loading of Actinide Chloride Surrogates in an Iron Phosphate Glass
Authors: Schofield James M. , Bingham Paul A. , Hand Russell J. 
Published: April 2011

AbstractPyrochemical and aqueous processing of impure Pu metal and Pu oxide results in a waste stream that contains calcium, plutonium and americium chlorides (CaCl2, PuCl3 and AmCll3) that tend to form insoluble salt layers on top of melts during vitrification. By adding chlorides to ammonium dihydrogen phosphate, ammonium chloride is evolved during vitrification, leaving behind the Ca, Pu and Am cations as oxides. HfO2 and Sm2O3 were used as surrogates for PuO2 and Am2O3 respectively. The effects of waste loading and melt duration on some basic physical properties for different melt durations were investigated. Synthetic batches containing CaO, HfO2 and Sm2O3 in molar ratios 93.34 : 5.91 : 0.74 respectively, reflecting the waste composition, were mixed with a fixed glass composition of P2O5 (75 mol%) and Fe2O3 (25 mol%). NH4H2PO4 was used as a precursor material. The solubility of calcium, hafnium and samarium oxides into phosphate glasses is reported and elemental analysis of the glasses is compared with expected glass compositions.
Mechanistic Investigation of Internal Corrosion in Nuclear Waste Containers Over Extended Time Periods
Authors: Onumonu Elsie , Stevens Dr Nicholas P.C. 
Published: April 2011

AbstractStorage of the UK's Intermediate Level Wastes (ILW), which comprises Magnox fuel cladding, uranium and small items of equipment exposed to radiation, is currently achieved via encapsulation within cementitious grout housed in 500 litre 316L stainless steel drums. The cements used display a high pH; in such an environment many metals form surface hydroxides or oxides. Magnox reacts with free water at high pH with the liberation of hydrogen whilst undergoing corrosion to form hydroxide species.Corrosion of Magnox cladding has previously been monitored by measuring the rate of hydrogen evolution and/or weight loss. Recent work by our group has shown impedance techniques may also be useful in monitoring early corrosion behaviour. In this project electrochemical polarisation techniques will be employed to examine the corrosion behaviour of Magnox fuel in situations where it is in electrical contact with other metals, including uranium, and hence determine how galvanic effects influence corrosion behaviour. In this paper we describe the background to such experiments along with some preliminary results.
Effect of Gallium Oxide on Phase Assemblage in Apatite and Whitlockite Hosts for Waste Immobilization
Authors: Gerrard Lee. A. , Fong Shirley. K. , Metcalfe Brian. L. , Donald Ian. W. 
Published: April 2011

AbstractTo immobilize halide and actinide ions present in specific ILW waste a process has been developed that uses mineral phases as the host material. The mechanism of substitution of gallium into these phases will have a large effect on the phase assemblage. This will inevitably affect the total amount of halide that can be immobilized in to total phase mixture.The full simulated waste stream composition containing varying concentrations (1–40 wt.%) of gallium oxide was studied. Also nominal compositions for gallium doped fluorapatites (Ca10-1.5xGax)F2(PO4)6 (x = 0, 0.25, 0.5, 0.75, 1.0) and gallium doped whitlockites Ca9Gay(PO4)6+y (x = 0.2, 0.4, 0.6, 0.8, 1.0) were prepared at 750–1050 °C.These were studied by powder x-ray diffraction (XRD) to determine the phase assemblage and solid solution limits of gallium in the apatite and whitlockite phases. It was found that a complete solid solution was formed between whitlockite, Ca3(PO4)2, and Ca9Gay(PO4)6+y. In the nominal apatite compositions it was found that gallium did not substitute into the apatite structure but was instead partitioned over Ca9Gay(PO4)6+y, gallium phosphate, and unreacted gallium oxide. At higher temperatures gallium suppressed the formation of the apatite phase and was largely partitioned into the Ca9Gay(PO4)6+y phase whereas at lower temperature the majority was present as unreacted Ga2O3. In the full DCHP compositions it was found that gallium is likely to be partitioned over a number of phases including apatite, cationdoped whitlockite and gallium phosphate.
Alpha-Decay Damage in Murataite-Based Ceramics
Authors: Stefanovsky S.V. , Lukinykh A.N. , Tomilin S.V. , Lizin A.A. , Yudintsev S.V. 
Published: April 2011

AbstractSamples of murataite ceramics with the composition (wt.%) 3.8 Al2O3, 10.5 CaO, 54.0 TiO2, 10.6 MnO, 6.0 Fe2O3, 4.6 ZrO2, 8.1 ThO2, 2.4 Cm2O3 (1.8 244Cm) and a specific activity of 5.5×1010 Bq/g were prepared by cold pressing and sintering at 1250 °C for 24 hrs or by melting and recrystallisation in a resistive furnace at 1325 °C and 1350 °C for 1 hr. In the sintered ceramics murataite polytypes with five-fold (5C) or three-fold (3C) repeats of the fluorite unit cell and crichtonite were found to be the major phases. Perovskite, pseudobrookite, and pyrochlore were observed as minor phases. The 5C polytype was rendered X-ray amorphous at a cumulative dose of 2.73×1018 μ–decays/g (0.21 dpa) whilst the 3C polytype, which contained only traces of Cm, remains crystalline at this dose. In the melted ceramics the 5C and 8C murataite polytypes were found to be the major phases (80–90 % of the bulk) and minor amounts of rutile, crichtonite and perovskite were also observed. Complete amorphization of the murataite polytypes in the ceramics melted at 1325 and 1350 °C was achieved at doses of 2.46×1018 μ-decays/g (0.19 dpa) and 2.53×1018 μ-decays/g (0.20 dpa), respectively.
The Use of Surrogates in Waste Immobilization Studies: A Case Study of Plutonium
Authors: Bingham Paul A. , Hand Russell J. , Stennett Martin C. , Hyatt Neil C. , Harrison Mike T. 
Published: April 2011

AbstractSurrogates are widely used in the research and development of nuclear wasteforms, providing detailed insight into the chemical and physical behaviour of the wasteform whilst avoiding the widespread (restricted and costly) use of radiotoxic elements in the laboratory. However, caution must be exercised when dealing with surrogates since no single element or compound perfectly mimics all aspects of the behaviour of another. In this paper we present a broad discussion of the use of surrogates in waste immobilization, drawing upon and highlighting our research into glass and ceramic wasteforms for the immobilization of bulk PuO2.
Durability of Borosilicate Glass Compositions for the Immobilisation of the UK's Separated Plutonium Stocks
Authors: Harrison Mike T. , Scales Charlie R. 
Published: April 2011

AbstractSeveral glass compositions are currently under investigation for immobilisation of the separated PuO2 that has been produced as a result of civil nuclear fuel reprocessing in the UK. Whilst a final decision on the fate of what ultimately will be over 100 tonnes of plutonium has yet to be made, all options for the disposition of this material are currently being investigated by Nexia Solutions Ltd on behalf of the Nuclear Decommissioning Authority (NDA).As one of the immobilisation options, vitrification in borosilicate glass could potentially provide a criticality-safe and stable waste form with durability suitable for long term storage and subsequent repository disposal. From an initial experimental survey of potential candidates, three borosilicate compositions were selected for a more detailed study of the waste loading and chemical durability: lanthanide borosilicate (LaBS), alkali tin silicate (ATS) and high-lanthanide alkali borosilicate (modified-MW). In these inactive tests, hafnium was used as the surrogate for plutonium. This paper describes a range of static leach tests that were undertaken in order to understand the overall durability of the waste forms, as well as the release rates of the Pusurrogate when compared to any neutrons poisons present in the glass. For the LaBS compositions it was found that the release rate of gadolinium was potentially slightly higher than that of hafnium, although both were as low as 10-5 to 10-6 g m2 day-1. The potential implications for long-term repository behaviour are discussed.
Waste Package Corrosion Studies Using Small Mockup Experiments
Authors: Anderson B.E. , Helean K.B. , Bryan C.R. , Brady P.V. , Ewing R.C. 
Published: April 2011

AbstractUnderstanding the corrosion of spent nuclear fuel (SNF) and the subsequent mobilization of released radionuclides, particularly under oxidizing conditions, is one of the key issues in evaluating the long-term performance of a nuclear waste repository. However, the large amounts of iron in the metal waste package may create locally reducing conditions that would lower corrosion rates for the SNF, as well as reduce the solubility of some key radionuclides, e.g., Tc and Np. In order to investigate the interactions among SNF-waste package-fluids, four smallscale (∼1:40 by length) waste package mockups were constructed using metals similar to those proposed for use in waste packages at the proposed repository at Yucca Mountain. Each mockup experiment differed with respect to water input, exposure to the atmosphere, and temperature. Simulated Yucca Mountain process water (YMPW) was injected into three of the mockups at a rate of 200 μL per day for five days a week using a calibrated needle syringe. The YMPW was prepared by equilibrating 50 mg/L silica as sodium metasilicate with air, and adding enough HCl to lower the pH to 7.6 in contact with an excess of powdered calcite.X-ray powder diffraction and scanning electron microscopy confirm that, where corrosion occurred, the dominant corrosion product in all cases was magnetite. In the high temperature (60°C) experiment, hematite and a fibrous, Fe-O-Cl phase were also identified. The Fe(II)/Fe(III) ratios measured in the corrosion products using a wet chemistry technique indicate extremely low oxygen fugacities (10-36 bar). Experiments are in progress in which 0.1g powdered UO2 was included in the mock-up in order to investigate the relative kinetics of Fe and U oxidation and to identify the U corrosion products formed under these conditions.
A New Method for Real-Time Monitoring of Grout Spread Through Fractured Rocks
Authors: Henderson Alasdair E. , Robertson Iain A. , Whitfield John M , Garrard Graham F.G. , Swannell Nicholas G. , Fisch Hansruedi 
Published: April 2011

AbstractReducing water ingress into the Shaft at Dounreay is essential for the success of future intermediate level waste (ILW) recovery using the dry retrieval method. The reduction is being realised by forming an engineered barrier of ultrafine cementitious grout injected into the fractured rock surrounding the Shaft. Grout penetration of 6m in <50μm fractures is being reliably achieved, with a pattern of repeated injections ultimately reducing rock mass permeability by up to three orders of magnitude.An extensive field trials period, involving over 200 grout mix designs and the construction of a full scale demonstration barrier, has yielded several new field techniques that improve the quality and reliability of cementitious grout injection for engineered barriers.In particular, a new method has been developed for tracking in real-time the spread of ultrafine cementitious grout through fractured rock and relating the injection characteristics to barrier design. Fieldwork by the multi-disciplinary international team included developing the injection and real-time monitoring techniques, pre- and post injection hydro-geological testing to quantify the magnitude and extent of changes in rock mass permeability, and correlation of grout spread with injection parameters to inform the main works grouting programme.
Release of U, REE and Th From Palmottu Granite
Authors: Markovaara-Koivisto Mira K. , Marcos Nuria , Read David , Lindberg Antero , SiitariKauppi Marja , Loukola-Ruskeeniemi Kirsti P. 
Published: April 2011

AbstractInterpretation of trace metal mobility in geological environments is often hampered by conflicting data from alternative experimental protocols and the lack of detailed mineralogical characterization of the host medium. To illustrate this issue, the release of uranium, thorium and the rare earth elements (REE) was investigated in polished rock slab samples from the U-Th deposit at the Palmottu Natural Analogue study site (SW Finland) by means of leaching experiments. The samples were sequentially leached with artificial groundwater of moderately high carbonate content at pH8, and nitric acid solutions at pH5 and pH3. The mineralogy and composition of the U, Th and REE mineral phases was studied using SEM-WDS and EDAX methods before and after each leaching step. In parallel, leaching was carried out on crushed material of the same samples and the leachates analysed by ICP-MS.The most notable U minerals are uraninite, uranophane and two secondary U-Pb phases. Thorium occurs predominantly in monazite and at lower concentrations in uraninite. Accessory thorite is also present, which together with monazite contains most of the REE. Differential leaching of the elements was noted across all phases on the timescale of the experiments. Uraninite is partly dissolved at pH3. The main secondary uranium phase, uranophane, was stable in moderately acidic solution, but easily dissolved in the artificial groundwater and at pH3. Some release of REE was observed although the main REE-bearing phase, monazite, showed no evidence of degradation.This study provides insights in the preferential release of radionuclides in granitic bedrock. An understanding of these processes is essential when assessing the safety of a spent fuel repository. Once released from the primary waste form U is expected to precipitate as secondary phases within micro fractures, as observed at Palmottu and numerous other deposits.
Inner Material Requirements and Candidates Screening for Spent Fuel Disposal Canister
Authors: Puig Francesc , Dies Javier , Sevilla Manuel , Pablo Joan de , Pueyo Juan José , Miralles Lourdes , Martínez-Esparza Aurora 
Published: April 2011

AbstractIn the context of the present Spanish ‘once-through’ nuclear fuel cycle, the need arises to complete the geological repository reference concept with a spent fuel canister final design. One of the main issues in its design is selecting the inner material to be placed inside the canister, between the steel walls and the spent fuel assemblies. The primary purpose of this material will be to avoid the possibility of a criticality event once the canister walls have been finally breached by corrosion and the spent fuel is flooded with groundwater. That is an important role because the increase in heat generation from such an event would act against spent fuel stability and compromise bentonite barrier functions, negatively affecting overall repository performance. To prevent this possibility a detailed set of requirements for a material to fulfil this role in the repository environment have been devised and presented in this paper. With these requirements in view, eight potentially interesting candidates were selected and evaluated: cast iron or steel, borosilicate glass, spinel, depleted uranium, dehydrated zeolites, haematite, phosphates, and olivine. Among these, the first four materials or material families are found promising for this application. In addition, other relevant non-performance-related aspects of candidate materials, which could help on decision making, are also considered and evaluated.
A Model for Predicting the Temperature Distribution Around Radioactive Waste Containers in Very Deep Geological Boreholes
Authors: Travis Karl P. , McTaggart Neil A. , Gibb Fergus G. F. , Burley David 
Published: April 2011

AbstractWe present a mathematical model for determining the temperature field around radioactive waste containers in very deep geological boreholes. The model is first used to predict the temperature rise for some simple, but well-established cases with known solutions in order to verify the numerical work. The temperature distribution is then determined for two variants of the deep bore hole concept; a low temperature variant and a high temperature variant. The results from these studies are discussed in terms of their utility in establishing deep borehole disposal as a workable concept.